JAEA-Review 2008-061
Annual Report of Fusion Research and Development Directorate of JAEA from April 1, 2007 to March 31, 2008
Fusion Research and Development Directorate
Japan Atomic Energy Agency Naka-shi, Ibaraki-ken
(Received October 20, 2008)
This annual report provides an overview of major results and progress on research and development (R&D) activities at Fusion Research and Development Directorate of Japan Atomic Energy Agency (JAEA) from April 1, 2007 to March 31, 2008, including those performed in collaboration with other directorates of JAEA, research institutes, and universities. The JT-60U operation regime was extended toward the long sustainment of high normalized beta (βN) with good confinement (βN = 2.6 × 28 s). Effectiveness of real-time control of current profile was demonstrated in high β plasmas. Toroidal momentum diffusivity and the convection velocity were systematically clarified for the first time, and intrinsic rotation due to pressure gradient was discovered. Effects of toroidal rotation and magnetic field ripple on type I ELM size and pedestal performance were clarified, and type I ELM control was demonstrated by toroidal rotation control. Variety of inter-machine experiments, such as JT-60U & JET, and domestic collaborations were performed. In theoretical and analytical researches, for the NEXT (Numerical Experiment of Tokamak) project, numerical simulations of a tokamak plasma turbulence progressed and a zonal field generation was investigated. Also, nonlinear MHD simulations found the Alfven resonance effects on the evolution of magnetic islands driven by externally applied perturbations. Integrations of several kinds of element codes progressed in the integrated transport/MHD model, the integrated edge/pedestal model and the integrated SOL/divertor model. In fusion reactor technologies, R&Ds for ITER and fusion DEMO plants have been carried out. For ITER, a steady state operation of the 170GHz gyrotron up to 800 s with 1 MW was demonstrated. Also extracted beam current of the neutral beam injector has been extended to 320 mA at 796 keV. In the ITER Test Blanket Module (TBM), designs and R&Ds on Water and Helium Cooled Solid Breeder TBMs were progressed. For the ITER TBM fabrication technology, a full-scale TBM first wall was fabricated with reduced-activation-ferritic-martensitic steel (F82H) by a hot isostatic press method. Tritium processing technology for breeding blankets and neutronics integral experiments with a blanket mockup were also progressed. For ITER and DEMO blankets, studies on neutron irradiation effects and ion irradiation effects on F82H steel characteristics were continued using HFIR, TIARA and so on. ITER Agreements entered into force on 24th October 2007. On the same day ITER Organization (IO) was established and JAEA was designated as the Japanese Domestic Agency (JADA) by Japanese Government. The Procurement Arrangement for the Toroidal Field (TF) Conductors was concluded between the IO and JADA in Nov. 2007, and then the contract with Japanese companies to fabricate TF conductors was launched in March 2008. The Quality Assurance Program of JADA inevitable to implement the procurement was approved by the IO. The Project Management (e.g. Schedule Management, Procurement Management, and QA) of JADA was executed. The preparation of the procurement was continued for the TF coil, Blanket First Wall, Divertor, Remote maintenance System of Blanket, EC and NB Heating System, and Diagnostics. The Agreement for the Broader Approach (BA) Activities entered into force and JAEA was assigned as the Implementing Agency for the BA Activities, on 1st June 2007. Contracts for the constructions of the buildings for the IFERC Project and the IFMIF/EVEDA Project as well as the preparation for the Rokkasho site were made in March 2008. For the IFERC Project, information exchange to obtain common view of DEMO design ______Editors: Kubo, H., Hoshino, K., Isei, N., Nakamura, H., Sato, S., Shimada, K., Sugie, T.
was carried out. Procurement Arrangement for the urgent tasks to be implemented for DEMO R&D was concluded. Preparation toward selection of a super computer for the Computational Simulation Centre started. For the IFMIF/EVEDA Project, detailed planning for the initial engineering research of the accelerator components, the lithium target and the test cell were implemented. On JT-60SA project, Executive Summary of the Conceptual Design Report including the baseline project schedule was approved at the 1st BA Steering Committee Meeting. The Integrated Project Team, consisting of the Project Team, JAEA Home Team and EU Home Team, was organized and developed the Integrated Design. The Procurement Arrangements were launched between the Implementing Agencies, JAEA and F4E, for the supply of PF magnet conductor and winding building, vacuum vessel and materials of in-vessel components. Finally, as to fusion reactor design study, physics design on current drive and beta limit analysis and neutronics design on the blanket for SlimCS progressed.
Keywords; Fusion Plasma, Fusion Technology, JT-60, JT-60SA, ITER, Broader Approach, IFMIF/EVEDA, International Fusion Energy Research Center, DEMO Reactor
Contents
I. JT-60 Program 1. Experimental Results and Analyses 1.1 Extended Plasma Regimes 1.2 Heat, Particle and Rotation Transport 1.3 MHD Instabilities and Control 1.4 H-Mode and Pedestal Research 1.5 Divertor/SOL Plasmas and Plasma-Wall Interaction 2. Operation and Machine Improvements 2.1 Tokamak Machine 2.2 Control System 2.3 Power Supply System 2.4 Neutral Beam Injection System 2.5 Radio-Frequency Heating System 2.6 Diagnostics Systems 2.7 Safety Assessment 3. Domestic and International Collaborations 3.1 Domestic Collaboration 3.2 International Collaboration
II. Theory, Simulation and Modeling 1. Numerical Experiment of Tokamak (NEXT) 1.1 Magnetohydrodynamic (MHD) Theory and Simulation 1.2 Plasma Turbulence Simulation 2. Integrated Modeling 2.1 MHD Stability -Effect of Equilibrium Properties on the Structure of the Edge MHD Modes in Tokamaks- - 2.2 SOL-Divertor 2.3 ELM Transport -Effect of Radial Transport Loss on the Asymmetry of ELM Heat Flux- 2.4 Heating and Current Drive -Electron Cyclotron Current Drive in Magnetic Islands of Neo-classical Tearing Mode- 2.5 Integrated Simulation 3. Atomic and Molecular Data
III. Fusion Reactor Design Study 1. Progress in Compact DEMO Reactor Study 2. Numerical Study on Beta Limit in Low Aspect RatioTokamak
Appendix A.1 Publication List A.2 Organization of Fusion Research and Development Directorate A.3 Personnel Data
I. JT-60 Program 1. Experimental Results and Analyses The JT-60U tokamak project has focused on the physics and engineering issues for the establishment of burning plasma operation in ITER and steady-state high β operation toward JT-60SA and DEMO. Since the optimization of current profile and pressure profile is essential for the stable and steady plasma operation, enhancing the control capability of these parameters is quite important. In addition, plasma rotation has a critical influence on magneto-hydro-dynamic (MHD) stability in the high β plasmas, edge pedestal Fig. I.1.1-1 (a) Waveforms of a discharge in which βN=2.6 performance and edge localized mode (ELM) was sustained for 28 s (and HH98y2≥1 for 25 s). (b) Extended operation regime toward longer sustainment characteristics. Thus, to expand the flexibility of heating and higher βNHH98y2 (filled circles) from previous and rotation profile control, the power supply system for campaigns (open circles). The inset is the q profile at
3 neutral beams (NBs) has been upgraded so as to t=27 s. Sustained duration at βN=2.6 has been almost enable the maximum injection period of 30 s in 2007. tripled. By making full use of the state-of-the-art tools of about 16 times of the current diffusion time (τR). Good feedback control, heating/current drive and diagnostics, confinement (HH98y2≥1) is kept for 25 s, where the significant progress has been made in the integrated period is limited by the degradation of confinement research on steady-state operation, transport, MHD (after t=29 s in Fig. I.1.1-1(a)) due to increase in density stability, and edge pedestal physics . caused by enhanced recycling. In this discharge, off-axis bootstrap current maintains the flat safety factor 1.1 Extended Plasma Regimes profile at qmin~1, and the values of bootstrap current Operation regime was extended toward the long fraction (fBS) and non-inductively driven current (fCD) sustainment of high with good confinement [1.1-1]. βN are 0.43 and 0.48, respectively. Weak and infrequent Real-time control of current profile was applied to high sawtooth activities were observed, but they did not plasmas, and the effectiveness of controlling the β affect the confinement. The onset of the neoclassical minimum in the q profile (q ) in high plasmas was min β tearing modes (NTMs) was successfully prevented by clearly demonstrated [1.1-2]. Driven current profiles for the optimization of pressure gradient at low-q rational the off-axis tangential NBIs were directly evaluated surfaces (3/2 and 2) through adjustment of the heating using motional Stark effect (MSE) diagnostics, and profiles. compared with ACCOME calculation [1.1-2].
1.1.2 Real Time Current Profile Control 1.1.1 Long Sustainment of High βN Plasma The minimum value of the safety factor profile (qmin) Toward the development of the ITER hybrid operation affects the MHD instability related to the low-q rational scenario where the large current (low q ) operation for 95 surface, such as m/n=3/2 and 2/1 NTMs, where m and n a long period is required both with high and a βN are the poloidal and toroidal mode numbers, moderate current drive (CD) fraction, JT-60U has respectively. In order to demonstrate the effectiveness extended the operation regime, making full use of the of qmin control on elimination of MHD activities, the increased heating power for a long pulse in this real-time control of qmin was applied to the high-βp campaign; see Fig. I.1.1-1. Power supply systems for 3 mode plasmas. Figure I.1.1-2 shows the waveforms of units of NBs (about 6 MW) are modified to enable 30 s the discharge at Ip=0.8 MA, Bt=2.4 T (q95=5.4). Neutral operation (formerly 10 s). The sustained period with beams (NBs) of about 14 MW were injected to produce βN=2.6 has been almost tripled from 10 s to 28 s in the high β plasma. When the diamagnetic stored energy high-βp ELMy H-mode plasma at low q95=3.2 Wdia reached 1.55 MJ (βN=1.7, βp=1.5), the m/n=2/1 (Ip=0.9 MA, Bt=1.5 T). The duration corresponds to
profile was measured for the first time. The total amount of the measured driven current agreed with calculations by the ACCOME code in both cases. In addition, the measured driven current profile was consistent withỞ neutron-emission profile measurement representing beam ion profile. However, the measured driven current density profile was more off-axis than that in the calculations.
References 1.1-1 Ide, S., the JT-60 team, Proc. 35th EPS Conf. on Plasma Physics, ECA 32F (2008) I1.007. 1.1-2 Suzuki, T., et al., Nucl. Fusion 48, 045002 (2008).
Fig. I.1.1-2 (a) NB injection power (dotted), LH injection 1.2 Heat, Particle and Rotation Transport power (thin solid), and its command value (thick solid). 1.2.1 Toroidal Momentum Transport and Rotation (b) qmin (solid) and its reference (dotted, qmin,ref=1.7). Gray Profile in L-mode plasmas line indicates qmin=2. (c) stored energy. (d) spectrogram of magnetic fluctuations. The control of qmin started from t=8 s. Recent tokamak studies have been emphasizing that the plasma rotation profile plays essential roles in NTM appeared, leading to decrease in Wdia by 22 %. In this discharge, we expected appearance of the m/n=3/2 determining confinement and stability. In order to establish a method for controlling the rotation profile, NTM, so we set reference qmin,ref=1.7 intending to eliminate q=1.5 rational surface in the plasma. The construction of physics basis of momentum transport lower hybrid (LH) waves were injected after t=7.5 s. and its effect on rotation profiles is required. Concerning toroidal momentum transport, most of the The qmin control by LH power (PLH) starts at t=8 s, with previous works evaluated the diffusive term utilizing the primary parallel-refractive-index N//~1.65. When the steady state momentum balance equation. However, it is qmin control started, PLH increased to raise qmin. The qmin reached to 1.7 at t=10 s, and the LH power decreased by recognized that the measured toroidal rotation velocity (V ) profiles cannot be explained by the momentum the control. The qmin overshot the reference qmin,ref, and t reached to 2 at t~11 s. At this timing, the magnetic and transport coefficient evaluated by the steady state density fluctuations at 1.6 kHz were suppressed, and equation. Parameter dependences of the toroidal momentum Wdia started increasing back to the initial value. Due to diffusivity ( ) and the convection velocity (V ), and the reduction of LH power from t~10 s, qmin decreased χφ conv the relation between and heat diffusivity ( ) are down to qmin,ref=1.7 at t~12 s, showing that LH current χφ χi systematically investigated in typical JT-60 L-mode drive was actually effective in sustaining the qmin value. plasmas using the transient analysis by using the 1.1.3 Off-Axis NBCD momentum source modulation [1.2-1]. Experiments Off-axis current drive is essentially important in have been carried out to investigate the momentum advanced operation in ITER and JT-60SA. Although transport. The absorbed power is varied from 2.4 to off-axis current drive by neutral beams (NBCD) is a 10.7 MW under otherwise similar conditions 3 candidate for the off-axis current driver, the (Ip=1.5 MA, Bt=3.8 T, q95=4.2, δ=0.3, Vp=74 m ). These characteristics of the driven current profile had not been plasmas stay in a low collisionality and small Larmor investigated yet due to the difficulty in measuring its radius regime with ρpol*~0.03-0.06 and ν*~0.07-0.26. broadly distributed current density profile. Using the As shown in Fig. I.1.2-1, the momentum diffusivity MSE diagnostics, the NBCD profile was measured in increases with increasing the heating power, and the shape of profile is nearly identical. The V profile plasmas with Ip=0.8 MA and 1.2 MA at Bt=3.8 T [1.1-2]. χφ conv In both cases, no MHD activity was observed except takes non-zero value and has a minimum value at ELMs, and the spatially localized NB driven current r/a~0.6. The momentum diffusivity at r/a=0.6 roughly
scaled linearly with the heating power in this data set. rotation. Although the progress in understanding the
We have also investigated Ip dependence of χφ and Vconv. physics of momentum transport and rotation has been
During this Ip scan (Ip=0.87, 1.5, 1.77 MA), one CO made experimentally and theoretically in worldwide, the tangential NB and one PERP-NB are injected with a characteristics of the rotation profile including the similar absorbed power Pabs=3.3-4 MW and other spontaneous term is not yet sufficiently understood. plasma parameters were Bt=3.8-4.1 T, δ=0.3 This is due mainly to the experimental difficulty in 19 -3 andne=1.3-1.9x10 m . The inverse χφ at r/a=0.6 separating the non-diffusive term and the spontaneous increases linearly with Ip, and such improvement of the term. momentum confinement is confirmed by steady toroidal We have identified the intrinsic rotation, which is momentum profiles. It is also found that toroidal not determined by the momentum transport coefficients rotation velocity profiles in steady state can be almost and the external momentum input [1.2-2]. The reproduced by χφ and Vconv estimated from the transient momentum transport coefficients, such as χφ and Vconv, momentum transport analysis at low β (βN<0.4) as can be obtained separately from the transient shown in Fig. 1.2-2. momentum transport analysis, and the Vt profile is calculated using these coefficients, the external torque
and the boundary condition of Vt [1.2-1]. From this approach, we have identified roles of externally induced
rotation and the intrinsic rotation on the measured Vt profiles. The heating power scan is performed both in
L-mode (Ip=1.5 MA, Bt=3.8 T, q95=4.2, δ~0.3,
κ~1.3-1.4) and in H-mode plasmas (Ip=1.2 MA,
Bt=2.8 T, δ~0.33, κ~1.4) in order to investigate the effect of plasma pressure on the spontaneous rotation. The absorbed power is varied over the range 2.4 MW
power scan in L-mode plasmas. (c) Dependence of χφ at discharges, and 4.8 MW r/a=0.6 on absorbed power. discharges. Although the measured Vt profile agrees with the calculation in the region 0.45 measured Vt deviates from the calculated one in the CTR-direction in the core region 0.2 in Fig. I.1.2-3. The difference in Vt is observed in the region where such large pressure gradients are measured. Fig. I.1.2-2 Experimental data (solid circles) and reproduced Vt profile (solid lines) in L-mode plasmas (a) with CO-NBI, and (b) with CTR-NBI, respectively. Torque profiles (dashed lines) are also shown. 1.2.2 Role of Pressure Gradient on Intrinsic Toroidal Rotation The toroidal rotation velocity profile is determined by the momentum transport, external momentum source and intrinsic plasma rotation. It is important to separately evaluate contributions of the intrinsic rotation Fig. I.1.2-3 (a) Profiles of the measured V (open circles) and the external induced rotation, and to understand the t and the calculated one (solid line), (b) and ∇Pi in the case mechanism responsible for the generation of intrinsic of higher βN=1.07 L-mode plasma with Pabs=11 MW. velocity on the physical processes determining the heat transport and the pedestal structure in H-modes is one of the key issues in recent tokamak research. In JT-60U and DIII-D, it has been observed that the energy confinement is improved with the toroidal torque (and the resulting rotation) in co-direction to the plasma current Ip by the tangential NB. The confinement improvement with co-NBI accompanies the enhanced plasma pressure at the top of the H-mode pedestal together with the reduction of ELM frequency in case of Fig. I.1.2-4 Difference between measured V and the t JT-60U [1.2-3]. However, an underlying physics calculated one (ΔVt) is plotted against the ∇Pi during heating power scan in (a) L-mode and (b) H-mode mechanism of this confinement improvement is not yet plasmas. clear. In this study, relation between heat transport in the A good correlation between the difference in Vt (i.e. plasma core and toroidal rotation as well as Vt(calculation) 㻐 Vt(measurement)) and the pressure characteristics of the pedestal structure were examined gradient is found during the heating power scan: ΔVt in conventional ELMy H-mode plasmas in JT-60U. increases with increasing pressure gradient in all cases Conducting the experiments on power scan with a including L-mode, H-mode, CO-, and CTR-rotating variety of toroidal momentum source generating the plasmas as shown in Fig. I.1.2-4. These results indicate plasma rotation direction to co, balanced and counter that the local pressure gradient plays the role of the local with respect to the plasma current, dependence of the value of spontaneous rotation velocity. heat transport properties in the plasma core on toroidal rotation profiles was investigated. Energy confinement 1.2.3. Dependence of Heat Transport on Toroidal improvement was observed with increase in the toroidal Rotation in Conventional H-Modes in JT-60U rotation in co-direction. Heat transport in the plasma Temperature gradients are a key element in driving core varies, while self-similar temperature profile in the turbulent convection and causing anomalous heat variations of toroidal rotation profiles are sustained. transport in plasmas. The property of the turbulence Pressure at the H-mode pedestal is increasing slightly driven by temperature gradient is believed to be with toroidal rotation in co-direction. Thus, energy provided by a strong increase of heat conduction which confinement enhanced with co-toroidal rotation is sustains a self-similar profile when the temperature determined by increased pedestal and reduced transport, profile exceeds a threshold in the temperature gradient brought on by profile resilience. In other words, heat (TG) scale length. The significant role of edge pedestal transport in the plasma core is mainly determined by the structure in H-modes, which is affected by the ELM saturation of temperature profile and is not strongly activities, has been observed in many devices as a influenced locally by toroidal rotation. As shown in boundary condition for the heat transport in the plasma Fig. I.1.2-5, when the pedestal temperature was fixed core. In JT-60U, the core temperatures vary in between the cases of co and counter-NBI by adjusting approximately proportion to the temperatures at the the plasma density, the identical temperature profiles shoulder of the H-mode pedestal in a wide range of were obtained in spite of totally different toroidal accessible densities in the type I ELMy H-mode regime, rotation profiles. In H-mode plasmas where the ion suggesting the existence of profile resilience. In H-mode channel is heated dominantly by the positive ion-based plasmas where the ion channel is heated sufficiently by neutral beams, the saturation of ion temperature gradient the positive ion-based neutral beams (NBs), the heat governs the heat transport in the plasma core. As a transport at the plasma core has been considered to be result, large increase in heat conduction imposes the imposed strongly by the existence of a critical scale resilient profile of ion temperature, under which local length of the ion temperature gradient LTi. effect of toroidal rotation profile on the scale length of Understanding the effects of toroidal rotation ion temperature gradient is very weak [1.2-4]. Fig. I.1.2-5 Profiles of (a) toroidal rotation, Vt, (b) electron density, ne, (c) ion temperature, Ti, (d) electron temperature, Te, (e) integrated ion conductive loss power, Qi and (f) ion heat diffusivity, χi for the cases of co and counter-NBI. The pedestal temperature is decreased by increasing density for the case of co-NBI. factor was defined by a ratio of the central electron 1.2.4 Comparisons of Density Profiles in JT-60U density at r/a=0.2 to the volume averaged density. In Tokamak and LHD Helical Plasmas with Low this figure, the abscissa indicates an electron-ion Collisionality collision frequency normalized by a trapped electron In order to understand particle transport systematically bounce frequency (ν*b). The normalized collisionality in toroidal plasmas, electron density profiles were of unity indicates a boundary between the collisionless compared in JT-60U tokamak and LHD helical plasmas region and the plateau region in both tokamak and with low collisionality. The neoclassical particle single helicity (where only a single helical Fourier transport in a low collisionality regime is significantly magnetic component exists) configurations. Thus, ν*b is different for tokamak and helical plasmas. In helical a good index for showing the collisionality range for plasmas, the 1/ν regime exists, where the neoclassical comparison. Here, the value of ν*b was calculated using transport is enhanced as being proportional to 1/ν (ν is plasma parameters at r/a=0.5. The density peaking collisionality) in non-axisymmetric helical plasmas due factor increases with decreasing ν*b in JT-60U. The to the presence of helical ripples [1.2-5]. On the other dependence in the collisionless region of LHD for hand, anomalous transport in both plasmas seems to be Rax=3.5 m tends to approach that in the collisionless related with common physics in toroidal systems. region of JT-60U, although the dependence in the Gyrokinetic analyses showed that the quasilinear collisionless region of LHD for Rax=3.6 m is reversed particle flux driven by drift wave instabilities exhibits compared with that in the collisionless region of weak dependence on the magnetic configurations JT-60U. [1.2-6]. The collisionality dependence of the density profile Figure I.1.2-6 shows dependence of density peaking in JT-60U could be related to the anomalous inward factors on collisionality for JT-60U ELMy H-mode pinch. The collisionality dependence of the density plasmas (Ip=1 MA and Bt=2 T) and LHD plasmas profile in LHD for Rax=3.5 m could involve the (Bt=2.8 T) with a magnetic axis (Rax) of 3.5 m and 3.6 m anomalous inward pinch, because it has been shown that [1.2-7]. Note that scaling studies indicated that both of neoclassical transport in LHD in the plateau region these plasmas have the same weak gyro-Bohm like becomes the less pronounced as the magnetic axis is confinement feature [1.2-8]. Here, the density peaking moved the more inward [1.2-5]. Therefore, the collisionality dependence of density profiles in LHD for 1.3 MHD Instabilities and Control Rax=3.5 m might become similar to that in tokamak 1.3.1 MHD Stability Analysis of the High β Plasma for plasmas, because gyrokinetic analyses showed that the the Resistive Wall Mode Experiments quasilinear particle flux driven by drift wave A very low rotation threshold was identified in JT-60U instabilities exhibits weak dependence on the magnetic experiments in an investigation of the critical rotation configurations [1.2-6]. On the other hand, the for stabilizing resistive wall mode (RWM) by changing collisionality dependence of density profiles in LHD for the toroidal plasma rotation using different combination Rax=3.6 m could be related to an increase in the of tangential neutral beams (NBs) [1.3-1]. Note that neoclassical outward flux, because the convective flux magnetic braking using external coils is not applied in for electrons is dominated by the temperature gradient JT-60U. The identified critical rotation is Vt ~20 km/s driven flux, which increases in outward direction with and corresponds to 0.3 % of the Alfvén velocity at the decreasing collisionality in the 1/ν regime, in the q=2 surface, which is much smaller than the previous datasets used here. prediction in DIII-D with magnetic braking. It is important for the RWM experiment to check the stability limit of external kink-ballooning mode without wall (no-wall limit) and with wall (ideal-wall limit) because RWM is destabilized between the ideal-wall limit and the no-wall limit. The MHD stability analysis code MARG-2D has been developed and has a useful interface for the analysis of experimental results with realistic plasma parameters, e.g. plasma pressure and current profile and equilibrium [1.3-2]. By using the MARG-2D code, the no-wall limit and ideal-wall limit in the RWM experiments was evaluated, and the value of the normalized beta (βN) at the no-wall limit was βN=2.3. Since the experimentally achieved βN was about 2.8, it is sustained above the no-wall limit. Figure I.1.3-1 shows the plasma Fig. I.1.2-6 Dependence of the density peaking factor as a displacement of n=1 RWM for the plasma at the no-wall function of an electron-ion collisionality normalized by a limit. Here, n is the toroidal mode number, and the trapped electron bounce frequency. Circles show data for displacement profiles with the poloidal mode number JT-60U ELMy H-mode plasmas. Open and closed squares from 1 to 6 are shown in this figure. This mode has a show data for LHD plasmas for Rax = 3.5 m and 3.6 m, respectively. global structure that contains internal and external components. Namely the peaks of m=2~4 components correspond to rational surfaces q=2~4. The components References 1.2-1 Yoshida, M., et al., Nucl. Fusion 47, 856 (2007). 1.2-2 Yoshida, M., et al., Physical Review Letters 100,105002(2008). 1.2-3 Urano, H., et al., Nucl. Fusion 47, 706 (2007). 1.2-4 Urano, H., et al., Nucl. Fusion 48, 085007 (2008). 1.2-5 Murakami S., et al., Nucl. Fusion 42, L19 (2002). 1.2-6 Yamagishi O., et al., Phys. Plasmas 14, 012505 (2007). 1.2-7 Takenaga H., et al., Nucl. Fusion 48, 075004 (2008). 1.2-8 Yamada H. et al., Nucl. Fusion 45, 1684 (2005). Fig. I.1.3-1 Plasma displacement of n=1 RWM for the plasma at βN=2.3 without wall. of m=1 and m>4 are non-resonant and externally resonant ones. Moreover, the amplitude of this mode at the low-field side is larger than that at the high-field side. This is consistent with the experimental measurements. By this numerical analysis, we can estimate the no-wall and ideal-wall limit using experimentally obtained profiles. By comparing with the numerical and experimental results, it is found that the experimentally obtained critical rotation is still low as βN increases toward the ideal wall limit. These results indicate that for large plasmas such as in future fusion reactors with low rotation, the requirement of the additional feedback control system for stabilizing RWM Fig. I.1.3-3 The region for complete stabilization in the is much reduced. space of EC driven current and ECCD deposition width. The solid curve is the boundary for the complete stabilization. 1.3.2 Simulation of Neoclassical Tearing Mode Stabilization In particular, effects of ECCD location and ECCD In JT-60U experiments, active stabilization of deposition width on stabilization were investigated in neoclassical tearing mode with m/n=3/2 or 2/1 using detail [1.3-3,4]. Figure I.1.3-2 shows magnetic island electron cyclotron current drive (ECCD) has been width of m/n=2/1 NTM for various ECCD location extensively performed. In addition, simulation of the (ρEC) and EC-driven current (IEC). The value of the stabilization of NTMs was performed using the TOPICS vertical axis is normalized by the island width before code combined with the modified Rutherford equation. ECCD. For the reference case, IEC=12 kA, the values of In the simulation, the coefficients in the modified ECCD deposition width, full width of saturated island Rutherford equation were determined by comparing and mode location are 0.08, 0.11 and 0.59 in the with experimental data. The TOPICS simulations were normalized minor radius, respectively. The ratio of found to well reproduce the behavior of NTMs in EC-driven current density to bootstrap current density at JT-60U experiments. By using these coefficients, the mode rational surface, jEC/jBS, is about 1. This prediction analysis on NTM stabilization was performed. condition is similar to that in JT-60U experiments. In this case, NTM can be completely stabilized when misalignment is within about half of the full island width. When the misalignment is comparable to the island width, NTM is destabilized, and the width becomes wider by about 20 %. Such destabilization is actually observed in JT-60U experiments. When EC wave power is doubled (~23 kA), the V-shape profile becomes wider, that is, larger misalignment becomes acceptable for complete stabilization. However, at the same time, the destabilization effect also increases. When EC wave power is further increased to 29 kA, while allowed misalignment does not increase, the destabilization effect further increases. This suggests that precise adjustment is required to avoid NTM destabilization and subsequent mode locking even when Fig. I.1.3-2 Magnetic island width after ECCD for various abundant EC wave power is available. In addition to ECCD location and EC-driven current. The value of ECCD location, ECCD deposition width is another vertical axis is normalized by island width before ECCD. factor affecting NTM stabilization. In general, narrower ECCD deposition width has stronger stabilization effect. large tokamaks [1.3-6]. In this experiment, the first Figure I.1.3-3 shows a region of complete stabilization measurement of the toroidal wave numbers of those in the space of IEC and δEC*. Here, δEC* is full width at spontaneously excited waves is described. The modes half maximum of ECCD deposition profile normalized due to D-ions have zero or small toroidal wave number by saturated island width. As shown in this figure, the kz. On the other hand, the measurement of finite kz in the boundary for complete stabilization increases roughly modes due to FP-ions supports the excitation of the 0.5 with IEC . Thus, stabilization effect is roughly Alfvén waves is the possible origin of FP-ICEs. 2 proportional to IEC/δEC , showing that ECCD width is an Figure I.1.-3-4 shows a typical intensity plot of the important factor for NTM stabilization. The dependence fluctuation amplitude as a function of time and is similar to the previously obtained result on m/n=3/2 frequency. The time sequence of NBs is indicated in NTM stabilization. the figure. Two sharp peaks, of which frequencies are corresponding to the fundamental and 2nd harmonic 1.3.3 Instability in Ion Cyclotron Frequency Range cyclotron frequencies of 3He near the outer midplane [1.3-5] edge of the plasma, appear when the tangential The fluctuations in ion cyclotron range of frequency positive-ion-based NBs (P-NBs) are injected. After (ICRF) are driven by the presence of non-thermal ion perpendicular P-NBs are injected, relatively broad peaks distribution in magnetically confined plasmas. ICRF due to D-ions are detected. A peak with the lowest antennas are used as pickup loops for detecting frequency below 10 MHz appears after negative-ion electrostatic and/or electromagnetic fluctuations. Two based NB is injected. This is considered to be due to sets of ICRF antennas, which are installed with the T-ions. It is also observed that the excited modes due distance of 1.67 m in the toroidal direction, are used in to FP-ions (3He and T-ions) have different this experiment and the toroidal wave number can be characteristics: driven by different neutral beams and evaluated due to the small pitch angle of the toroidal having different parameter dependence. ICE due to magnetic field line. Two types of magnetic fluctuations T-ions has no harmonics and the value of ω/Ωci is are detected: one is due to high energy D-ions from smaller than that due to 3He-ions. Both beam-driven neutral beam (NB) injections and the other is due to ICEs and FP-ICEs are clearly observed, and those fusion products (FPs) of 3He and T-ions. These spatial structures are also obtained on JT-60U. fluctuations have been reported as ion cyclotron emissions (ICEs) in the burning plasma experiments on 1.3.4 Development of Active MHD Diagonostic System [1.3-7] In order to actively diagonose the MHD stability such as RWM, edge localized mode (ELM) and Alfvén eigenmode (AE), we have been developing an active MHD diagnostic system in JT-60U. This system has two one-turn coils located 180º apart toroidally in the JT-60U vacuum vessel. These coils are connected to bipolar power supplies that can apply ±60 V and ±60 A as the maximum voltage and current, respectively. According to the calculation by Biot-Savart method in vacuum, the maximum magnitude of radial magnetic perturbations is predicted to be δBr ~ 0.1 G at the plasma surface. Figure I.1.3-5 shows the Fourier components spectrum of magnetic perturbations produced by this Fig. I.1.3-4 Typical intensity plot of the fluctuations in the system. In the case where the coil currents are out of case of tangential and perpendicular NBs. Two types of phase and in phase with each other, the dominant mode fluctuations (due to injected D-ions and fusion product ions of 3He and T) are shown. The toroidal magnetic field number is odd and even, respectively. These toroidal strength is 2.3 T in this discharge. mode spectra are fairly broadened up to the intermediate JET, on the other hand, δr can be varied by selecting the appropriate differential current between odd and even set of coils out of 32 TF coils, providing in this case four levels of ripple amplitude of δr=0.1, 0.5, 0.75 and 1 %. Although the same level of ripple amplitude was successfully obtained in both devices, the Vt in JET was still higher than that in JT-60U due to a different level of fast ion losses. Since a correlation between δr and VT was found in JET plasmas, it is difficult to separate the two parameters, δr and Vt, in JET experiments. A series of power and density scans indicated that plasmas with lower δr and/or larger co-Vt are favorable Fig. I.1.3-5 Fourier spectra of magnetic perturbations produced by the active MHD diagnostic system. (a) Odd and (b) Even connections. n~20. This enables us to perform the active MHD diagnostic for low-n (RWM and low-n AE) to intermediate-n (ELM and AE) MHD instabilities. We expect that these could lead to deeper understanding of MHD stability. References 1.3-1 Takechi, M., et al., Phys. Rev. Lett. 98, 055002 (2007). 1.3-2 Aiba, N., et al., Comput. Phys. Commun. 175, 269 (2006). 1.3-3 Isayama, A., et al., Nucl. Fusion 47, 773 (2007). 1.3-4 Ozeki, T., et al., Phys. Plasmas 14, 056114 (2007). 1.3-5 Ichimura, M., et al., Nucl. Fusion 48, 035012 (2008). 1.3-6 Dendy R.O., et al., Nuclear Fusion 35, 1733-42 (1995). 1.3-7 Matsunaga, G., et al., Proc. 10th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems 2007 (Kloster Seeon, 2007) CD-ROM file P-10. 1.4 H-Mode and Pedestal Research 1.4.1 Effect of Toroidal Field Ripple and Toroidal Rotation on H-Mode Performance and ELM Characteristics in JET/JT-60U Similarity Experiments In order to understand the effect of toroidal field (TF) ripple and toroidal rotation (Vt) on H-mode performance and ELM characteristics, dedicated ripple experiments Fig. I.1.4-1 Comparison of electron density and were performed in JET and JT-60U using a matched temperature at the top of pedestal in ripple scan plasma shape [1.4-1]. After the installation of ferritic experiments in JT-60U (a) and in JET (b). ELM ped 19 -3 characteristics changed at ne ~2.4×10 m from type I steel tiles in JT-60U, toroidal field ripple, δr, near the ELMs to type I+III ELMs in JT-60U. outer midplane was reduced to ~0.5 % at Bt =2.2 T. In ped to achieve higher p and HH98y2 factor in both devices. spatial width of the H-mode edge transport barrier on As for ELM characteristics, larger co-Vt seems to local and global plasma parameters is not well increase the ELM energy loss together with the understood. In particular, knowledge of the pedestal ped reduction of the ELM frequency. Sustaining a high ne width Δped based on dimensionless parameters is of great with type I ELMs was still difficult in JT-60U plasmas help for the extrapolation towards a next step device. ped even with FSTs (δr ~ 0.5 %), while ne in JET plasmas However a well-validated means of predicting the at δr = 0.75 % could be increased to similar value as at δr spatial width of the H-mode transport barrier in ITER is = 0.1 % using the same amount of gas puffing, as shown still lacking and remains a major topic of research. in Fig. I.1.4-1. Therefore, there are still some Attempts to project a pedestal width for ITER through differences between the two devices. In order to obtain a physics based or empirical scaling from multi-tokamak better prediction for ITER plasmas, further databases are underway to extend turbulent models investigations into the effects of the TF ripple and the through the pedestal region in recent tokamak research. toroidal rotation are necessary, and the acceptable level A variety of empirical scalings of Δped during the of ripple amplitude in ITER is still uncertain from these well-developed type I ELMy H-mode phase have been experimental results in both devices. proposed empirically, based on theoretical or semi-empirical models. However, these scalings vary 1.4.2 Dimensionless Parameter Dependence of from machine to machine and with the operational H-mode Pedestal Width Using Hydrogen and regime.Ở In particular, a main argument on the pedestal Deuterium Plasmas in JT-60U width Δped is whether Δped scales as the normalized In H-mode plasmas, the edge pedestal structure poloidal Larmor radius ρpol* or βpol. This disagreement determines the boundary condition of the heat transport of the scalings intermingling ρpol* and βpol can be of the plasma core. Therefore, it is of primary caused by the existing edge stability boundary for ELMs. importance to understand the physical processes In this region, ρpol* and βpol are linked strongly and thus determining the edge pedestal structure. The structure of are hard to separate out in the standard dimensionless the H-mode pedestal is composed of a height of the parameter scan using a single gas species. To plasma pressure and a spatial width in which a steep distinguish these variables, a pair of experiments in pressure gradient is formed. In this region, the periodic hydrogen and deuterium plasmas was conducted in this expulsion of energy and particles is commonly observed study. Explicit difference between ρpol* and βpol is the 0.5 due to the existence of the edge localized modes mass dependence of ρpol* (∝ m ) in contrast with no (ELMs) triggered by a steep pressure gradient and/or a mass dependence in βpol. large bootstrap current. However, the dependence of the Fig. I.1.4-2 (a) Pedestal Ti profiles for the deuterium (shaded circles) and hydrogen plasmas (open circles) at the same βpol and ν*. (b) Data area in βpol – ρ*pol space of the experiments. (c) Relation between ρ*pol and Δped/a. The exponent of ρ*pol in the log-linear regression at fixed βpol is ξ = 0.2. Closed and open inverse triangles indicate the deuterium and hydrogen plasmas at a given βpol (~0.23), respectively. Both the database analysis and the dedicated experiments on the mass scan indicated that the pedestal (a) Line-averaged ) 4 8 width depends very weakly on the plasma particle -3 electron density ( CO 2 ) m species or ρpol*. Identical profiles of the edge Ti 19 ( MW ) ( 10 2 4 obtained in the experiments suggested that the pedestal width more strongly depended on βpol than ρpol* as 0 0 shown in Fig. I.1.4-2. The experiment on scan was βpol 2.0 (b) Particles also performed. Higher βpol plasma had higher pedestal 1.0 Ti value accompanied by wider pedestal width in spite 23 atoms) x10 of almost identical ρpol* at the pedestal. Based on the 23 0.0 ( 10 experiments on the dimensionless parameter scan, the -1.0 scaling of the pedestal width was evaluated as Δped ∝ ap 0.2 0.5 -2.0 ρpol* βpol [1.4-2]. 45950 45955 45960 45965 45970 45975 Pulse number References 1.4-1 Oyama, N., et al., “Effect of toroidal field ripple and Fig.I.1.5-1 (a) line averaged electron density and (b) toroidal rotation on H-mode performance and ELM particle retention as a function of shot number. characteristics in JET/JT-60U similarity experiments,” to be published to Journal of Physics: Conference 1.5.2 Radiative Plasma Control Series. Reduction of heat loading appropriate for the plasma 1.4-2 Urano, H., et al., Nucl. Fusion 48, 045008 (2008). facing components is crucial for a fusion reactor. Power 1.5 Divertor/SOL Plasmas and Plasma-Wall handling by large radiation power loss using impurity Interaction (Argon) gas seeding was studied to sustain the high 1.5.1 Particle Control in Long-Pulse Discharges confinement ELMy H-mode plasma with the large The particle retention in the plasma surface wall has radiation power under the wall saturated condition, been investigated in a series of long-pulse discharges by where particle recycling flux changes during the long a particle balance analysis [1.5-1]. As shown in discharge. Controllability of the large radiation in the Fig.I.1.5-1, the first 13 discharges with a line-averaged good energy confinement plasma (HH98y2=0.78-0.87) electron density lower than 3.5×1019 m-3, which was investigated by the radiation feedback control of the corresponds to ~75 % of the Greenwald density, the Ar gas puff rate, using the bolometer signals viewing particle retention in the first wall gradually decreased. the main plasma edge. Total radiation fraction of This is due predominantly to outgassing from the Prad/Pabs = 0.75-0.9 was maintained continuously during divertor plates with an increase of the surface Ar gas puffing (up to 13 s so far) under the outgassing temperature. In contrast, in the high-density discharges condition from the plasma facing components. with a line-averaged electron density of 75 – 83 % of the Greenwald density, the particle retention changed to 1.5.3 SOL Fluctuations and ELM increase since the wall-pumping rate becomes positive. Characteristics of the ELM filaments and intermittent Several reasons are candidates for the positive plasma events have been investigated from the fast wall-pumping rate and investigated. First, with sampling database of three reciprocating Mach probes increasing electron density, the increase of the divertor measured in 2006. In particular, in-out asymmetry in the plate temperature was suppressed, resulting in lower plasma propagation of these transient events was outgassing rate compared to that of the low-density investigated [1.5-2, 3]. discharges. Second, in the high-density discharges, the ELM plasma measurements showed two transient co-deposition of deuterium with carbon became convective fluxes during the ELM event: one was the significant to enhance the fuel retention rate. Third, in large, short convective flow caused by the filaments, addition, wall-pumping process was enhanced and the other was the flow reversal over a wide SOL dynamically with increasing particle flux. region. The latter fact was observed only at the Fig.I.1.5-2 Enlarged time evolutions of ELM event: (a) Dα brightness in the main plasma and the HFS divertor, (b) HFS js at upstream and downstream sides of the HFS Mach HFS probe. Delay of the first js peak, separation and duration HFS HFS HFS of these peaks are shown by τ (peak), τp-p and δtpk . High-Field-Side (HFS) SOL. During the early period of type I ELM event as shown in Fig. I.1.5-2, seven large peaks in the ion saturation current (j ) appeared only at s Fig.I.1.5-3 (a) Time evolution of j (b) conditional averaging HFS-up s the upstream side of the HFS SOL (js ), and the results of js and floating potential Vf at LFS SOL. radial propagation was not seen. On the other hand, at the Low-Field-Side (LFS) midplane, multi-peaks in js were seen both at the upstream and downstream sides of decay. This feature was similar to that of theoretical the Mach probe, and radial propagation with prediction for plasma blobs. The conditional average of v⊥=0.4-3 km/s was evaluated. In Fig. I.1.5-2, delays of floating potential (Vf) measured simultaneously showed the first few peaks were smaller than the parallel that Vf changes from positive to negative with respect to convective transport from the LFS midplane (~190 µs). 〈Vf〉 as a large burst of js passes by. This result gives the This fact suggested that ELM filaments extend to the internal potential structure of plasma blobs. The radial HFS edge and are ejected into the SOL. In the velocity of the plasma blob was estimated by using characteristics of the filaments at the HFS SOL, the delayed time between the peaks of js and Vf to be about mid separation of the maxima (τp-p ) was 25-55 µs, which 0.43 km/s. was larger than those at the LFS midplane. The toroidal mode number was estimated to be relatively large 1.5.4 Spectroscopic Study in Divertor (n=18–44) from the toroidal rotation velocity The spectral lines of C2+ and C3+ emitted around the measurement. After the appearance of the multi-peaks, X-point in a detached plasma with MARFE were the flow reversal of the SOL plasma occurred over a measured with a VUV spectrometer and a wide SOL region. two-dimensional visible spectrometer in order to Intermittent convective plasma transport, i.e. understand the controllability of the dominant radiation “plasma blobs” has been observed in the LFS SOL, from C3+ [1.5-4]. As shown in Fig. I.1.5-4, it has been which is thought to play a key role for cross-field found that C3+ was produced by the volume transport of the SOL plasma. Conditional averaging recombination of C4+ and the ionization of C2+ method was employed to reveal the typical burst’s comparably. In contrast, the volume recombination of profile, and Figure I.1.5-3 shows the typical time C3+ was not detected, and the ionization flux of C3+ was 3+ 3+ evolution of js at the LFS SOL and the positive spikes less than 1 % of the C generation flux. Thus, the C have the common property of a rapid increase and slow generation flux was higher by two orders of magnitude than the loss flux. This result suggests that loss the toroidal movement (0.2-0.5 km/s) was faster than mechanism of large number of C3+ ions from the that of the radial movement (<0.05 km/s). Toroidal radiation dominant region such as parallel or cross-field movement was mostly towards the ion drift direction transport is significant. Because the ionization flux of (Ip), which is consistent with the SOL flow C3+ was much smaller than the recombination flux of measurement at HFS and LFS SOLs. C4+ around the X-point, the predominant source of C4+, which recombined into C3+, was presumably the main 1.5.6 Progresses of SOL and Divertor Code plasma. Since the flux of C4+ is determined by the Development and Simulation Study transport in the main plasma, it is difficult to control the In order to understand the divertor pumping for particle C4+ flux. control, the roles of atomic and molecular processes of Similarly, significant ionization of C2+ into C3+ and ionization (Ie), charge exchange (CX), elastic collision no recombination of C3+ into C2+ indicated that the (EL), and dissociation (DS) were investigated by the source of C2+ exists in the divertor region, which can be SOLDOR/NEUT2D code, which coupled the 2D plasma increased for instance by seeding CD4. Because the fluid modeling code SOLDOR and the Monte Carlo source rate of C3+ from the main plasma (the neutral kinetic modeling code NEUT2D. Characteristics recombination of C4+) and the divertor (the ionization of of the incident neutrals into the exhaust slot were C2+) was found to be similar as described above, the investigated from the dependence of the neutral radiation loss control by impurity seeding from the pumping efficiency on the strike point distance from the slot. A case of small distance (2 cm) showed high pumping efficiency, where the most of neutrals produced on the target go toward the exhaust slot directly, except for the disappeared neutrals by Ie and DS processes. Whereas, for the long distance case of 10 cm, those neutrals were scattered at random by CX and EL and a few of them go toward the slot, leading a low pumping efficiency. Atomic and molecular processes was also identified when the divertor plasma ave 19 -3 changed from low ne /high Te (ne-sp =0.7×10 m and ave ave 19 -3 Te-sp =60 eV) to high ne /low Te (ne-sp =21×10 m Fig.I.1.5-4 From the left, the radiation from ionizing ave and Te-sp ≤1 eV). The incident neutrals to the slot for component, the ionization flux, the spatial distribution of the low n /high T case were dominated by atoms the emissivity, the recombination flux, and the radiation e e from recombining component of C3+ ( upper ) and C2+ produced through DS, then CX and EL. Molecules due ( lower ). to EL were are dominant for the high ne /low Te case [1.5-5]. divertor will work partly. A kinetic neutral model and a fluid neutral model in the divertor code have been compared under the 1.5.5 Study of Dust Dynamics detachment condition through collaboration with KEIO Movement and velocity of the dusts were measured with University. The SOL radial profiles of basic plasma and a fast visible TV camera (2-6 kHz) from tangential port neutral parameters with the fluid neutral model were to determine the trajectory and velocity of the dusts. In fitted to those with the kinetic model as the first step of the main SOL, many dusts with various directions were the comparison, where the drift effects were also observed, particularly, in the first shot after hard introduced. In the SOL, no significant effect of the drifts disruptions and overnight (3-7 hours) He-GDC. Many on the radial profiles of the plasma temperature and dusts were ejected from the inner divertor, in particular, density was seen. However, in the divertor region, large for the high strike-point case: large ELM heat and differences were seen in the two models without the particle loading on thick deposition layers may enhance effects of drift, and drift effects tended to enlarge the producing dusts due to large thermal stress. Velocity of difference. As a result, improvement of the fluid neutral model is required in the divertor region. [1.5-6]. effect of the plasma flow along magnetic fields on the tungsten transport was also investigated. 1.5.7 Tungsten Erosion and Deposition Tungsten tiles have been installed at the upper row of 1.5.8. Hydrogen-Isotope Retention the outer divertor in the No.8 port section since 2003. At the outer board first wall graphite tiles, it has been Poloidal and toroidal distributions of tungsten found that the deuterium retention was higher than the deposition on the divertor CFC/graphite tiles were hydrogen retention even after hydrogen discharges for investigated. A neutron activation method was used for tritium degassing in 2004 [1.5-7]: at the surface of the the absolute measurement of W density by the tiles (< 0.1 µm), the hydrogen retention was higher than 186W(n,γ)187W reaction by slow neutrons in JAEA/FNS the deuterium retention. In contrast, the deuterium (Fusion Neutronics Source). Conventional surface retention was higher in the deeper range between analysis methods such as EDX (Energy Dispersive 0.1 µm and 1 µm (corresponding to the implantation X-ray spectrometry) and XPS (X-ray photoelectron depth of the fast ion), which contributed largely to the spectroscopy) were also used. total amount of hydrogen-isotope retention, although it For the divertor tiles exposed in 2003-2004, decreased with the depth. In a further depth range tungsten deposition on the dome tiles was found only between 1 µm and 10 µm, the deuterium retention near the top surface (within depth of a few µm), while decreased gradually and finally reached a very low ratio tungsten on the inner divertor tiles was co-deposited of D to C (D/C~10-4) [1.5-8]. with carbon to the depth up to about 60 µm. The The hydrogen-isotope retention in the side surfaces neutron activation method could measure tungsten in of the inner and the outer board first wall tiles was these thick co-deposition layers. Ion beam analysis such measured, and D/C was found to be comparable to that as PIXE (Particle Induced X-ray Emission) was not of the outer divertor tiles [1.5-9]. Compared to the tile appropriate for this case since only near-surface surface of the inner divertor tiles, where the tungsten (~10 µm) was precisely measured by this hydrogen-isotope retention was due predominantly to method. It was found that tungsten deposition profile the co-deposition with carbon, D/C in the side surfaces was not uniform toroidally. Tungsten surface densities of the first wall tiles was smaller. Since the total area of near the inner strike point and on the outer wing were the tile side surface of the first wall tiles is much larger much higher at the toroidal position of 0 degree (near than that of the inner divertor tiles, the contribution of the W-tile position) than those toroidally separated at the tile side is under investigation. The deuterium depth 60 degree (~3 m). Here the toroidal angle was defined profile was similar to that of boron, indicating that the as the angle viewed from the top of the torus deuterium was incorporated with boron. (counter-clockwise). The deuterium depth profile of the divertor tiles Detailed measurement of the W toroidal distribution without the degassing discharges at the end of 2005-6 on the outer wing showed significant localization near operations was measured for the first time [1.5-10]. the W-tile section. Since ionization length of W atom in Comparison of the deuterium depth profile at the inner the divertor plasma was more than an order of divertor tile surface with and without the degassing magnitude shorter than the distance of ~3 m, this discharges showed that about 90 % of the trapped long-distance deposition on the outer wing was not deuterium was removed by the hydrogen degassing attributed to the local deposition of sputtered tungsten discharges. However, in the range deeper than ~1 µm, atoms. This result indicated that transport of tungsten the deuterium retentions with and without degassing ions in the outer divertor plasma was significantly discharge were comparable within a factor of 2 (except affected by the cross-field transport in the private flux for the dome top tile). Furthermore, it has been found region. Toroidally localized W deposition was also that the deuterium concentration in the deep area (up to observed on the inner divertor tile, thus significant 16.4 µm) was not decreased. A similar depth profile was amount of the tungsten could be also transported to the also found in the divertor tiles of ASDEX-U, and it was inner divertor through the private plasma by the concluded that this is due to permeation of hydrogen cross-field transport such as drifts. At the same time, isotopes along the orientation of carbon fibers [1.5-11]. This is also a candidate mechanism to explain the wall-pumping in the series of the high-density and long-pulse discharges [1.5-1]. 2. Operational and Machine Improvements 2.1 Tokamak Machine References 2.1.1 Development of Supersonic Molecular Beam 1.5-1 Nakano, T., et al., Nucl. Fusion 48, 085002 (2008). 1.5-2 Asakura, N., et al., “ELM propagation in the low- and Injection high-field-side Scrape-off Layer of the JT-60U In order to control plasma pressure at the pedestal tokamak,” to be published in J. Phys.: Conf. Ser. region and to evaluate the effect of fuel on the (2008). self-organization structure of plasma, a supersonic 1.5-3 Asakura, N., et al., “Application of Statistical Analysis to the SOL Plasma Fluctuation in JT-60U,” submitted molecular beam injection (SMBI) system has been to J. Nucl. Mater. installed in JT-60U device. The SMBI system has been 1.5-4 Nakano, T., et al., “Radiation process of carbon ions collaborated with CEA/Cadarache in France. The in JT-60U detached divertor plasmas,” submitted to J. schematic diagram of the SMBI is shown in Fig. I.2.1-1. Nucl. Mater. 1.5-5 Kawashima, H., et al., Contrib. Plasma Phys. 48, 158 The nozzle heads were put on the wall of vacuum vessel (2008). (VV). 1.5-6 Hoshino, K., et al., Contrib. Plasma Phys. 48, 136 (2008). 1.5-7 Yoshida, M. et al., “Hydrogen isotope retention in the first wall tiles of JT-60U,” submitted to J. Nucl. Mater.. 1.5-8 Hayashi, T. et al., J Nucl. Mater. 363-365, 904 (2007). 1.5-9 Nobuta, Y. et al., “Retention and depth profile of hydrogen isotopes in gaps of the first wall in JT-60U,” submitted to J. Nucl. Mater. 1.5-10 Hayashi, T. et al., “Deuterium depth profiling in graphite tiles not exposed to hydrogen discharges Fig. I.2.1-1 Schematic diagram of the SMBI system. before air ventilation of JT-60U,” submitted to J. Nucl. Mater. 1.5-11 Roth, J. et al., J Nucl. Mater. 363-365, 822 (2007). Fig. I.2.1-2 Time evolution of plasma parameters during the gas-jet injection using SMBI. However, in the first SMBI system, the seal of the nozzle head was evaporated during baking of the vacuum vessel at a temperature of 300 degree C, causing a large seal leak. To improve this, heat-resistant materials were tested at JAEA and CEA/Cadarache, and FLUORITZ-HR was confirmed as a most promising seal material that showed no evaporation at temperature lower than 280 degree C. Thus, FLUORITZ-HR was adopted as a new seal material, and the baking temperature was decreased from 300 degree C to 280 degree C and gas pressure of the nozzle was increased from 0 MPa to 0.5 MPa to reduce the possibility of leak Position of broken-away tile due to deformation of the seal. After these countermeasures, the SMBI successfully worked without vacuum leak. Time evolutions of the plasma parameters in the case of high field side injection of SMBI are shown in Fig. I.2.1-2. The frequency of the SMBI is 10 Hz, and the pressure of SMBI is 0.3 MPa. The fast response of the plasma density synchronizing with the movement of Fig. I.2.1-3 Photograph of circumstance of the broken-away the nozzle head of SMBI is confirmed. tile. 2.1.2 Break-away of the Carbon Tiles 2.2 Control System A carbon tile in the outer dome of the divertor was 2.2.1 New Real-Time Control Functions for Advanced broken away and the tile dropped into the vertical port Plasma Operation in June, 2006. The cause of the break was considered to For efficient exploration toward high performance be due to the heat flow from a plasma to the carbon tiles. plasmas in JT-60 operation, more advanced real-time The trace of heat concentration was observed at the bolt control functions are required to be applied to the and nut between the tile and the basement of tile. The experiment. The major developments conducted in 2007 nut was melted by the heat flow and the tile was broken are presented as follows: (a) New real-time control of away from the basement. ion temperature (Ti) profile and toroidal rotation Countermeasures for this break-away were taken velocity (Vt) measured by the fast charge exchange during the maintenance period from May to September recombination spectroscopy (CXRS) system during in 2007. In the inspection just after the break-away, bolt neutral beam heating [2.2-1]. (b) The integral control holes with unprocessed edges, which made a slight gap method newly employed for the real-time control of between the tile and the basement to reduce heat radiation power profile in a divertor plasma. (c) A conduction between them, were found commonly for digital filter newly applied to the separatrix strike some of the dome tiles. Therefore, all bolt holes of the position on the divertor dome plates for stable strike dome basements were processed. In addition, carbon point control. (d) Multiple integral controls applied to a sheets were set between the dome tiles and basements to single real-time profile control. (e) Prompt plasma shape ensure heat conduction though the dome had originally evolution viewer newly developed for efficient been designed as components that were not expected to evaluation before the creation of JT-60 database using directly receive high heat flux. However, the the real-time plasma shape visualization system [2.2-2, break-away of the carbon tile in the outer dome was 3]. (f) Voltage control newly developed for safe found in March, 2008 again. The position of the tile is operation even in case of deteriorated electrical the same as that of the broken-away tile in June, 2006. insulation of the PF coil power supply. These new The circumstance of the broken-away tile is shown in advanced plasma real-time control functions have Fig. I.2.1-3. All tiles in outer dome (125 tiles) were contributed to achievement of annual mission checked again. No loosened bolt and melting were parameters specified in JT-60 experimental plan. found except the broken-away tile. It is considered that the countermeasures for the first break-away was 2.2.2 Development of an FPGA-based Timing Signal effective except the broken-away tile, although the Generator cause has not been understood perfectly. The same The timing system in the supervisory control system of countermeasures were taken again for the broken-away, JT-60 [2.2-4] is composed of specially ordered CAMAC and the position of the outer separatrix leg has been modules for limited usage [2.2-5]. This system controlled at more than 0.5 cm away from the dome generates trigger signals and a clock signal necessary since this event. for measurements and controls in the JT-60 experiment. Recently, we have been facing difficulties in shown in Fig. I.2.2-3. Test result is shown in Fig. I.2.2-4 cost-effective maintenance due to CAMAC [2.2-6]. The elapsed transmission time was 276.5-326.5 standardized hardware, and in increasing troubles due to ns, and reading time from input register was 50-100 ns, the aged deterioration. and writing time to output register was 62.5 ns. The The following design guidelines for a new system optical transmission time takes 333 ns between configuration were employed: (a) A master clock cycle supervisory timing system and subsystem. Total sum of should be changed from 1 kHz to 40 MHz to improve test result and optical transmission time take 600-650 ns accuracy. (b) FPGA (Field Programmable Gate Array) in signal transmission while 30-40 µs is needed in the was chosen to minimize delay time for the logical original CAMAC system. This value is less than calculation, and enhance the flexibility for changing allowable delay time 1 µs. logic flows. (c) The addition and change of the timing signal should be easily conducted in the new system. (d) Oscilloscope The timing signal should be transmitted to the subsystem in less than 1 µs by using the optical cable. (e) The number of cables should be minimized, which Function Generator requires multiplex transmission. (f) The interfaces Timing Optical fiber Timing A Signal cable (4 m) Signal B Generator Generator between the supervisory timing system and subsystem 1 2 one should remain unchanged. Fig. I.2.2-3 Configuration of timing signal transmission test. Timing system supervisor Timing Signal Generator Ethernet Slave Slave Master 1 N VME Crate Particle Supply A B Discharge and Heating Plasma Equilibrium Timing system Control System Control System Control System 276.5 ~ 326.5 ns host computer C Input/ Contents Time P Output U Optical Connector Connector Timing Read register 50~100 ns Signal Generator FPGA Logic operation ~0 ns Subsystem timing system 1 (118.75 ns) Parallel-Serial conversion 68.75 ns * Subsystem Component Optical Transfer (4 m) 14 ns * Timing Serial-Parallel conversion 81.25 ns * Signal Generator FPGA Logic operation ~0 ns 2 Fig. I.2.2-1 New timing system hardware configuration. (143.75 ns) Write register 62.5 ns ():Processing time, *: Calculated value Fig. I.2.2-4 Elapsed time chart in the timing signal Test Switch Clock transmission test of test results Serial-Parallel FPGA Converter (Top) Electric to / from Optical References Converter 2.2-1 Yoshida, M., et al., “Real-time measurement and feedback control of ion temperature profile and (Front) toroidal rotation in JT-60U,” submitted to Fusion Eng. Des. Status Input-Output Clock Connector JTAG Error Optical Power Indication Connector Connector Indication Connector Switch 2.2-2 Sueoka, M., et al., Fusion Eng. Des. 83, 283 (2008). 2.2-3 Sueoka, M., et al., Fusion Eng. Des. 82, 1008 (2007). Fig. I.2.2-2 Circuit configuration of the timing signal 2.2-4 Kurihara, K., et al., Fusion Eng. Des. 81, 1729 generator. (2006). 2.2-5 Akasaka, H., et al., Fusion Eng. Des. 71, 29 (2004). Figure I.2.2-1 shows a new timing system 2.2-6 Kawamata, Y., et al., Fusion Eng. Des.Ở 83, 198 hardware configuration. The timing system supervisor is (2008). composed of the VME-bus system with the host computer through Ethernet network. Figure I.2.2-2 2.3 Power Supply System shows the circuit configuration of the timing signal 2.3.1 Operational Experience generator. Response time measurement of timing signal Annual inspections and regular maintenances for the transmission has been carried out by using two timing power supply system have been conducted to maintain signal generators connected with optical fiber cable as high performance operation as shown in Table I.2.3-1. In addition, the renewal and special maintenances have A problem of the excessive shaft vibration was been also conducted to avoid troubles for the observed in the test operation with two-single-layer aged-deteriorated system as shown in Table I.2.3-2. connection. Voltages between three phases for two These activities contributed to achieving safe operation single layers of stator windings are shown in Fig. I.2.3-1 of the systems. (a) and (b). The voltages from the phase V commonly oscillated, while the voltage of U-W phases showed no Table I.2.3-1 Annual inspections and regular maintenances oscillation. of the power supply system. When two single layers were connected together, no Item Term The Toroidal Field Coil Power Supply April - October excessive shaft vibration occurred at all as shown in Fig. (TFPS) 2007 I.2.3-2. It could be understood that the V-phase stator The Poloidal Field Coil Power Supply April - October (PFPS) 2007 winding might be somehow differently reconstructed The Power Supply for Additional Heating September 2007 from the original one in 2004. Devices (H-MG) The Grounding Systems January - February 2008 The Lightning Arrester System January 2008 Vv-w Vu-v The Power Distribution Systems July 2007 1 cycle Table I.2.3-2 Renewals and special maintenances as Vw-u measures for the aged-deteriorated power supply system. Rotating cycle normalized by the pole number Item Term Fig. I.2.3-2 Output voltage ripple in case the two stator Inspection of metal enclosed switchgears July 2007 windings are connected together in parallel (Rotational (M/C) for the Power Distribution System -1 (every year) speed: 580 min , Exciter current: DC93 A). Renewal of MG shaft vibrograph for the January 2008 H-MG Maintenance of outside metal enclosed August - 2.4 Neutral Beam Injection System switchgears for the TFPS September 2008 In the campaign of 2007, the injection time of the Mass-spectroscopy of resolved gas in February 2008 insulation oil for step-down transformers negative-ion-based NBI (N-NBI) unit was successfully extended from previous 20 s to 27 s at 1.0 MW by 2.3.2 Investigation of MG Shaft Vibration in the modifying the negative ion sources. Four perpendicular Motor-Generator for Additional Heating Devices positive-ion-based NBI units were also upgraded to The motor-generator with 300-ton flywheel for extend the injection pulse length up to 30 s at 2 MW of additional heating devices (H-MG) has double layers of injection power. These long pulse injections from the stator windings connected together to a single output N-NBI and P-NBI units significantly contributed to the through the AC circuit breakers. study on quasi-steady state plasmas in JT-60U. 2.4.1 Long Pulse Operation of NBI System Vv-w Vu-v On JT-60U, there are 11 positive-ion-based NBI (P-NBI) units, each of which injects 2 MW D0 beams. 1 cycle Vw-u Out of P-NBI units, 4 tangential P-NBI units have been 0 Rotating cycle normalized by the pole number already upgraded to inject the D beams for 30 s in 2006. (a) Single Layer A In 2007, 4 perpendicular P-NBI units were additionally Vv-w Vu-v upgraded to extend the injection pulse length up to 30 s by mainly increasing the capacity of the power supplies. The injection pulse length of the upgraded P-NBI units 1 cycle Vw-u is extended while the beam-limiter temperature Rotating cycle normalized by the pole number increased by heat load of the high-energy re-ionized (b) Single Layer B 0 particles from the D beams was confirmed to be Fig. I.2.3-1 Output voltage ripple in the stator winding (Rotational speed: 580 min-1, Exciter current: DC93 A). reduced to an allowable level. The pulse lengths of the upgraded P-NBI units are extended from the previous 10 s to ~ 20 s. By using the tangential units modified in some of the beamlets were mis-deflected and struck on 2006 and the perpendicular units, high-power and the acceleration grids, resulting in the high grid power long-pulse D0 beams were reliably injected to meet the loading. To reduce the grid power loading, a thin plate requirements of plasma physics. This significantly for tuning the steering angle of the beamlets, called as contributed to the study on quasi-steady state plasmas in “field shaping plate (FSP)” was newly designed and JT-60U. tested. Figure I.2.4-1 shows the vertical deflection angle The injection pulse length in the N-NBI on JT-60U of the outmost beamlets as a function of thickness of the was also extended from the previous ~20 s to 27 s by FSP. To suppress the interceptions of the ion beam by optimizing the beam steering angle to reduce the grid the grids and the beam-limiters, a 1 mm thick FSP was power loading, as explained in 2.4.2. The time response chosen and installed on the JT-60U negative ion source. of the neutron flux from the plasma indicates constant The use of the FSP reduces the grid power loading D0 power without degradation during ~30 s. Since the normalized by the drain power from the previous value voltage holding capability of the ion sources was as of 7.5% to 6%, which is acceptable for the full poor as 290 keV due to insufficient conditioning time, specifications of the D- ion beam [2.4-2]. the D0 beam was injected from one ion source, and In the JT-60U negative ion source, the electrons are hence the D0 power was no more than 1 MW. Using two mainly deflected downwards by the stray magnetic field. negative ion sources, the long-pulse beam at the higher Some of the electrons are ejected from the ion source power is to be injected [2.4-1]. and dissipated on the inertial-cooled parts. Therefore, the ejected electron power should be quantified for long 2.4.2Ở R&D Programs of the Negative-Ion-based NBI pulse operation. To measure the electron power, a thin System for the Performance Improvement stainless steel plate was set below the beam path, and There remain three major issues to improve the the surface temperature of the plate was measured by performance of the negative-ion-based NBI system. One infra-red camera (Fig. I.2.4-2(a)). Figure I.2.4-2(b) is to reduce the grid power loading below an acceptable shows the typical temperature profile and the heat flux level. Second is to understand the electron heat load in on the plate when 300 keV and 3.4 A beam was the beamline. Third is to improve the voltage holding produced from only the central segment in the upper ion capability of the ion source, where the usable source. It is found that the highest heat flux from one acceleration voltage has been limited to < 400 kV. segment is < 8 W/cm2. Its total power load is ~2.6% of the drain power. For full beam with five segments, the highest heat flux is estimated to be ~37 W/cm2 by assuming the heat flux distributions for the other segments to be the same as that measured from the central segment. This power loading can be removed readily by inertia cooling even for 30 s long pulse operation [2.4-3]. The D0 beam power is restricted due to poor voltage holding capability of the ion sources. Even after sufficient conditioning, the achieved acceleration voltage is < 400 kV. Therefore, the voltage holding Fig. I.2.4-1 Beam steering by attaching a thin plate on capability should be increased to the rated value of 500 extraction grid. keV. As the first step for the improvement of the voltage holding capability, the breakdown location of The JT-60U negative ion source was originally the JT-60 negative ion source has been examined. It is designed to produce high current beams of 22 A at 500 found that the breakdown location varied with the keV for 10 s through 1080 apertures that are distributed conditioning stage. In the early stage, breakdown occurs on five segmented grids. The beamlets must be steered mainly in vacuum at gaps between the grids and their to focus the overall beam envelope. It was found that support frames with the total surface area of 2.5 m2. Careful observation shows that the conditioning Moreover in the ECH system, the world highest power gradually progresses with the breakdowns occurring at output of 1.5 MW for 1 s was achieved by a gyrotron to many different locations over the large surface area and the dummy load. Performance of the RF heating system with the reduction of the outgassing from the grids and has been constantly improved to extend the parameter the frames. This result suggests that the baking of the region of experiments. grids and frames would be effective to shorten the conditioning time. Over ~400 kV after conditioning of 2.5.1 Long-Pulse Operation of the ECH System several months, the breakdown location is changed to The extension of the pulse duration of the ECH system the surface of the FRP insulator with an inner diameter has been tried to enhance the plasma performance in the of 1.8 m. It is recently found in Saitama University that recent experiment campaign in JT-60U focusing on long the flashover voltage of a low outgassing epoxy resin sustainment of high performance plasmas. (~10-4 Pa·m/s) is twice higher than that of the Improvements of the vacuum pumping system of the conventional one. The use of the low outgassing epoxy transmission lines has been carried out in order to avoid resin is expected to improve the voltage holding pressure rise in the transmission lines due to capability [2.4-4, 5]. temperature rise of the components. Vacuum pumping unit was installed at the Matching Optics Unit (MOU) individually for each transmission line. By means of the techniques of controlling heater current and anode voltage during the pulse to keep the oscillation condition, pulse duration of 30 s at 0.4 MW (at gyrotron) has been achieved as shown in Fig. I.2.5-1. Temperature increase of the transmission line components shows availability of the present system for Fig. I.2.4-2(a) Set up of electron deposition measurement, 100 s operation in JT-60SA only for one shot. While (b) temperature and heat flux profiles at the electron dump. upgrade of the cooling system of waveguides will be required to repeat 100 s operations. References 2.4-1 Hanada, M., et al., Rev. Sci. Instrum. 79, 02A519 (2008). 2.4-2 Ikeda, Y., et al., “Recent R&D activities of negative ion based ion source for JT-60 SA,” to be published in IEEE Transactions on Plasma Science, 2008. 2.4-3 Kamada, M., et al., Rev. Sci. Instrum. 79, 02C114 (2008). 2.4-4 Hanada, M., et al., “Power loading of electrons ejected from the JT-60 negative ion source,” to be published in IEEE Transactions on Plasma Science, 2008. 2.4-5 Kobayashi, K., et al., “Conditioning characteristic of DC 500 kV large electro-static accelerator in Fig. I.2.5-1 Extension of pulse duration of ECH system. negative-ion-based NBI on JT-60U,” Proceedings of 23rd international symposium on Discharges and Electrical Insulation in Vacuum, Bucharest, 2008. 2.5.2 High Power Test of the Gyrotron Improvements of the ECH system (1 MW for 5 s per 2.5 Radio-Frequency Heating System unit) had been performed until FY2006 toward the The FY2007 was quite fruitful year for the JT-60U higher power and the longer pulse. A Si3N4 ceramic radio-frequency (RF) heating system. Pulse length DC-break was introduced into a gyrotron instead of an injected to the plasma by the electron cyclotron heating alumina ceramic DC-break due to the higher thermal (ECH) system reached 30 s of the middle term (-2009) strength. In order to avoid heating of a bellows which objective over the 25 s of the annual objective. enables to move the last mirror in the gyrotron, a cover was installed around the bellows to reflect stray-RF. Cooling water of a cavity wall was increased by about 20 %. Moreover, a high-power dummy load system (1.5 MW, CW) was developed to measure the oscillation power from the gyrotron. Those improvements enabled the ECH system to try high power oscillation up to 1.5 MW. In July of 2007, 1.5 MW for 1 s oscillation was achieved for the first time by means of fine optimization of the oscillation parameters as shown in Fig. I.2.5-2 [2.5-1]. The pulse length was not limited by any interlock signals. Therefore, it will be possible to obtain a longer pulse by making more fine adjustments. Fig. I.2.5-3 Power increase in LHRF system after repairing one of the damaged modules in the launcher. Reference 2.5-1 Kobayashi, T., et al., Plasma Fusion Res., 3, 014 (2008). 2.6 Diagnostics Systems 2.6.1 Real-Time Measurement and Feedback Control of Ion Temperature Profile and Toroidal Rotation Toward the steady-state operation with high beta and high thermal confinement, the real-time measurement and feedback control systems have been developed. Fig. I.2.5-2 Oscillation power versus pulse length achieved in the world (power of over 1 MW and pulse length of over 0.1 s). 2.5.3 Improvement and Performance of the LHRF System The LHRF (Lower Hybrid Range of Frequency) experiments such as real-time control of plasma current profile were performed. In FY 2006, 6-modules out of 8-modules of the launcher had been used because of Fig. I.2.6-1 Schematic diagram of CXRS system for damage at the launcher mouth due the malfunction of real-time feedback control. the arc detector in 2005. In the August of 2007, one of the heavily damaged mouths was repaired carefully from inside of the vacuum vessel of JT-60U, and the protection function against the arcing was upgraded with a new infrared camera watching at the launcher mouth from a tangential port. The conditioning operation for the launcher having 7-modules started in September and the injected power was reached ~ 2 MW in March 2009. The improvement of the power from the achievement of 1.6 MW in 2006 was 1.25 which was Fig. I.2.6-2 (a) Reference grad-Ti and measured grad-Ti. more than estimation (1.17~7/6) by a recovered module (b) Injected units of controlled NB heating. (c) Radial profiles of T with measurement point (solid symbols). as shown in Fig. I.2.5-3. i A fast charge exchange recombination spectroscopy are selected to make a beam focusing better [2.6-2]. (CXRS) system has been developed for the real-time Performances of the ion gun have been investigated on a measurement and feedback control of ion temperature test stand [2.6-3]. A beam current of 10 mA and a (Ti) profile and toroidal rotation velocity (Vt) [2.6-1]. In divergence angle of 0.2 degrees and equivalent current order to control Ti and Vt in real-time, the charge of 3 mA at the observation area are attained by the ion exchange recombination spectroscopy with high time gun. Figure I.2.6-4 shows the obtained current is resolution, the real-time processor system, and the increased with the extracted current. Moreover, a long real-time control system have been developed (Fig. I. pulse operation of 50 seconds with beam current of 10 2.6-1). Utilizing this system, real-time control of the Ti mA has been demonstrated. Then the ion gun has been gradient has been demonstrated with NBs at high beta installed and operated on JT-60U successfully. After plasmas (βN~1.6-2.8). The strength of the internal beam conditioning, a first signal of the lithium beam transport barrier is controlled (Fig. I. 2.6-2). Moreover, emission (2 2S - 2 2P resonance Ở line) has been the real-time control of Vt has been demonstrated from obtained in JT-60U plasma. Adjusting etalon filters, the counter to direction. Then the behavior of ELM changed lithium beam probe system is operated for the edge by controlling the Vt. current measurement. Fig. I.2.6-4 Experimental and simulation results of obtained current at 6.5 m. The saturation is caused by a shift of the neutralization position of an ion beam. 2.6.3 Polarization Interferometer for Thomson Scattering Recently, the use of a polarization interferometer based on Fourier transform spectroscopy has been proposed for Thomson scattering diagnostics [2.6-4]. It is possible that this method alleviates some of the disadvantages of Fig. I.2.6-3 Lithium beam probe system on JT-60U. Neutral conventional grating spectrometers. Furthermore, this lithium beam is injected to JT-60U via 6.5 m beam line. A method delivers a simple and compact system. We are beam monitor is installed in the middle of the beam line. developing the polarization interferometer for Thomson scattering diagnostics with YAG laser to demonstrate 2.6.2 Zeeman Polarimetry using Lithium Beam Probe the proof-of-principle. For thermal electrons, the optical for Edge Current Measurement coherence of the Thomson scattered light at an A lithium ion beam probe has been developed for edge appropriately chosen optical path delay, is a unique current measurement (Fig. I.2.6-3). A lithium ion gun function of T and n . The detection system utilizes a has been designed by the numerical simulation taking e e single bandpass filter combined with imaging optics and the space charge effects into account because a Zeeman dual detectors to simultaneously observe both dark and polarimetry requires low beam divergence angle. A bright scattered light interference fringes. The porous tungsten disk heated by an electron beam is normalized intensity difference between the bright and utilized for an ion emitter. The concave surface of the dark interference fringes gives a direct measure of T . A disk and a peaked heating profile of the electron beam e schematic of the polarization interferometer for (1) Surface Analyses of Ferritic Steel First Wall Tiles Thomson scattering diagnostics is shown in Fig. I.2.6-5. In FY2005, 1122 ferritic steel (8Cr-2W-0.2V) tiles were Scattered light is collected and introduced to the installed at the out board first wall inside the vacuum polarization interferometer through a fiber-optic bundle. vessel in order to reduce toroidal magnetic field (TF) This polarization interferometer consists of an objective ripples. For the surface analysis of these tiles removed lens as the fiber coupling optics, a band pass filter, a from the vessel, an application for use of radioisotopes polarizer, a birefringent plate which gives optical path induced on the tiles has been filed to the radiation delay, a Wollaston prism, an imaging optics to detector, regulation division of Ministry of Education, Culture, and dual APD (silicon avalanche photodiode) detectors. Sports, Science and Technology (MEXT) in March Since proof-of-principle tests will be carried out in 2007 and was subsequently permitted in April. TPE-RX using the existing YAG laser Thomson scattering system before experiments in JT-60U, (2) Extension of Neutral Beam Injection Time parameters for design of a prototype polarization Pulse lengths of neutral beam injectors (NBI) were interferometer are fixed as follows: Te ≤ 1 keV, ne ≥ extended from 30 s to 60 s in order to meet experiments 5×1019 m-3, scattering angle 90°. In an initial test using a with long pulse plasma discharges up to 60 s. An blackbody radiation source, the magnitude of the change application for the NBI pulse extension was permitted in in fringe visibility agrees with the numerical calculation. June 2007 and the experiments started in September This result confirms that, following suitable calibration, elongating the beam pulses. we will be able to sense visibility changes due to changes in the electron temperature. 2.7.2 Radiation Safety Assessment for JT-60 Decommissioning (1) Ở Radiational Exposure during Decommissioning Potential radiational exposure to workers in the torus hall during disassembly of the JT-60 was preliminarily assessed for internal and external radiation doses together with procedures of the disassembly. It was Fig. I.2.6-5 Schematic of a polarization interferometer for found that uses of the method of bubble lubricant during Thomson scattering diagnostics. diamond wire sawing and of some Green Room facilities for cutting the main structures including the References vacuum vessel sectors shall limit spread of 2.6-1 Yoshida, M., et al., “Real-time measurement and contamination and eliminate pollutant in the torus hall, feedback control of ion temperature profile and toroidal rotation using fast CXRS system in JT-60U,” greatly contributing to minimize their internal exposure. submitted to Fusion Eng. Des. In external radiation exposure, a contact dose level was 2.6-2 Kojima, A., et al., Plasma Fusion Res., 2, S1104 assessed less than 20 µSv/h inside the vacuum vessel at (2007). one year cooling time after a shutdown of JT-60, 2.6-3 Kojima, A., et al., “Development of a High-Brightness indicating an acceptable radiation environment for the and Low-Divergence Lithium Neutral Beam for a workers. The amount of internal and external exposure Zeeman Polarimetry on JT-60U,” to be published in doses for the workers was expected to be 100 times less Rev. Sci. Instrum. 2.6-4 Hatae, T., et al., Plasma Fusion Res., 2, S1026 (2007). than that of the regal limit, 1mSv/week inside the controlled area. 2.7 Safety Assessment 2.7.1 Application Works for Operational (2) Storage and Management for Radio-Activated Modifications on JT-60 Materials Removed from Controlled Areas Based on the law concerning prevention from radiation From both points of view of resources saving and hazards due to radioisotopes, etc., licensing procedures reusable nature of the materials, the radio-activated have been done regarding two items below. devices or structures replaced with new equipment are expected to be stored appropriately until the potential clearance rule is put into operation. It was found that content is 0.2 wt%, are used for the base plates of the some additional controlled on-site facilities or areas first wall inside the vacuum vessel. The major source of other than the existing JT-60 storage building are the activated level of the waste, therefore, takes about required for their storage and management. 45 years until less than the clearance level due to the long half-time nuclide 60Co. (3)㻃 Estimation of Low Level Waste by a Regulatory Clearance. 2.7.3㻃 Nuclear Shielding Assessment using ATTILA Low level waste of the JT-60U has been estimated by a Code regulatory clearance [2.7-1]. The JT-60U consists of Neutron fluxes of a tokamak fusion device with a main devices including a vacuum vessel, magnetic coils, cryostat were calculated with the three dimensional heating devices such as neutral beam injectors and radio (3D) nuclear shielding code, ATTILA. ATTILA is a frequency systems. Those structural materials include numerical modeling and simulation code designed to copper, stainless steel, carbon steel, high manganese solve the 3D multi-group Sn transport equations for steel, inconel 625, ferritic steel, lead and others. The neutrons, charged particles, and infrared radiation on an gross weight of the devices is about 6,400 tons. unstructured tetrahedral mesh. It uses a traditional Sn Neutron and gamma-ray fluxes during an operation source iteration technique for solving the first order were calculated with the ANISN code. Induced activity form for the transport equation. of the materials was calculated by the ACT-4 code in Figure I.2.7-2 depicts a demonstration result of the the THIDA-2 code system at various times after an total neutron flux distribution for a typical operational shutdown. The total neutron yields was superconducting tokamak with a cryostat.Ở From the assumed to be 1.14x1020 n for fourteen operation years results, ATTILA has proved to be useful to analyze the of the JT-60U. nuclear shielding properties especially for the port or Figure I.2.7-1 shows time evolutions of the volume duct streaming of an experimental building. of activated materials after the shutdown. The report of IAEA RS-G-1.7 [2.7-2] was referred to compare clearance levels to the results of the activated nuclides induced on the structural materials. The amount of the low level waste of which activation levels exceed the clearance levels is about 5000 tons just after the shutdown. On the other hand, the amount below the clearance levels is 1400 tons. Fig. I.2.7-2 Total neutron flux distribution of a typical superconducting tokamak device (by ATTILA code). 2.7.4㻃 Development of High Heat Resistant Neutron Shielding Resin Fig.I.2.7-1 Time evolutions of the volume of activated In fusion tokamak devices, temperature near a vacuum materials at the various cooling time after the shutdown. vessel is expected to rise up to ~300 oC, because the wall conditioning by vessel baking is of crucial The former decreases with time and vanishes at 45 importance for plasma discharge operations. years after the shutdown. Asymmetrically, the latter Shielding materials such as polyethylene and increases year by year as shown in the figure. In JT-60U, concrete are widely used. Polyethylene is the most stainless steel SS316 of about 50 tons, of which Cobalt popular resin for neutron shielding, but the heatproof sample. The glass transition temperature, an indicator of temperature is fairly low. Concrete is not suitable for the the heatproof temperature, in the specimen was obtained additional shielding material in the restricted space up to 320 oC. A test sample piece of the new resin around the center of the devices, while it endure a high before gel formation is shown in Fig. I.2.7-3. temperature environment. The heatproof neutron shielding resins such as KRAFTON-HB4 [2.7-3] and References EPONITE [2.7-4] had been developed by 2004. 2.7-1 Sukegawa, A. M., et al., “Estimation of Low Level Waste by a Regulatory Clearance in JT-60U Fusion KRAFTON-HB4 was the epoxy-based resin that Device,” 15th International Conference on Nuclear contains boron to reduce the production of secondary Engineering (ICONE-15) , April 22-26, 2007, Nagoya, gamma rays. It was developed for future FBR shielding JAPAN. materials. KRAFTON-HB [2.7-5] was improved one 2.7-2 IAEA, SAFETY GUIDE, No. RS-G-1.7. with the aim of suppressing the nuclear heating of 2.7-3 Ueki, K., et al., Nucl. Sci. and Eng., 124, 455 (1996). superconducting coils of the DD nuclear fusion devices. 2.7-4 Okuno,K., Radiation Protection Dosimetry, 115, 1-4, The maximum heatproof temperature of it was 150 oC. 258-261 (2005). 2.7-5 Morioka, A., et al., J. Nucl. Sci. Technol., Supplement EPONITE was the Colemanite and epoxy-based resin 4, 109-112 (2004). that contains boric acid. It was developed for a medical 2.7-6 Morioka, A., et al., J. Nucl. Mater., 367-370, 1085 equipment PET cyclotron with a heatproof temperature (2007). of 200 oC. 2.7-7 Sukegawa, A. M., et al., “High Heat Resistant Neutron Shielding Resin, ” 11th International Conference on Radiation Shielding (ICRS-11) , April 13-18, 2008 , Callaway Gardens, Pine Mountain, Georgia, US. 10 mm 200 mm 20 mm Fig. I.2.7-3 A test piece of the new epoxy-based neutron shielding resin before gel formation (Present study includes the result of “Research and Development for High Heat Resistant of Gel-type Neutron Shielding Resin“ entrusted to Japan Atomic Energy Agency by MEXT). Based on the previously demonstrated results above, we started the development of new resins against much higher temperature range up to 300 oC, and successfully produced the new boron-loaded and heat-resisting resin in 2005 [2.7-6]. The heatproof temperature has been improved by an appropriate mixing of stiffening materials with the epoxy-based resin. In 2007, the research and development of the Gel-type heat-resisting resin were initiated in response to the need for more flexible and light shielding materials expected to be useful in situations where an additional shielding is required in narrow or hard-to-reach areas such as locations of diagnostic collimators [2.7-7]. The results obtained to date have been limited to that of a solid-state 3. Domestic and International Collaborations In FY2007, 149 persons participated, who came 3.1 Domestic Collaboration from 19 research organs in Japan. Two leaders, one JT-60U was assigned as a core national device for joint from university or NIFS and the other from JAEA, research by the Nuclear Fusion Working Group of the occupy each subtheme. The number of research subjects Special Committee on Basic Issues of the Subdivision of the joint research was 27 in total in FY 2007, on Science in the Science Council of MEXT in January categories of which are shown in Table I.3.1-1. 2003. Using the JT-60 tokamak and other facilities, Twenty-two journal papers and 1 paper in conference JAEA has performed research collaboration with the proceedings were published as a result of the joint universities and National Institute for Fusion Science research in FY 2007. (NIFS). Accordingly, the joint experiments on JT-60 between JAEA and the universities by assigning 3.2 International Collaboration university professionals as leaders of research task Status of collaborative research based on the IEA forces has been successful. The number of collaborators Implementing Agreement on cooperation on the Large had increased significantly since FY 2003, and it is kept Tokamak Facilities is described first. The result of around 150 for recent five yeas as shown in Fig. I.3.1-1. personnel exchanges among Japan, US and Europe are as follows. The number of personnel exchanges, to which JAEA relates, is 11 in total. There are two personnel exchanges from JAEA to EU, 3 from EU to JAEA, 0 from JAEA to US, and 10 from US to JAEA. JT-60U contributes to ITPA/IEA inter-machine experiments. There are 3 to 7 experiments for each Topical Groups. In 2007, 4 related papers were published in journals and 10 related presentations are accepted for IAEA Fusion Energy Conference in 2008. As for remote collaboration, JAEA provides JT-60 data within the scope of the proposal document sheet (PDS). There are 8 active PDS (3 with EU, 2 with US, 2 with AUG, 1 with EAST). Fig.I.3.1-1 Evolution of number of collaborators in the Remote experimental system (RES) with high JT-60 joint research. network security has been developed in JT-60U. The remote experimental system is produced by personal authentication with a digital certificate and encryption Table I.3.1-1 Number of research subjects of the JT-60 joint of communication data to protect the JT-60U research according to category in FY 2007. supervisory control system against illegal access. No. of research Remote experiment in JT-60U was demonstrated from Category subjects Kyoto University (Japan) in 2006 and internationally Performance improvement 3 from IPP Garching (Germany) in 2007 on the occasion of an neo-classical tearing mode stabilisation similarity Transport 3 experiment between JT-60U and ASDEX Upgrade. Pedestal 2 RES was successfully verified on its authentication, MHD/High energy particles 7 encryption and turn around time. The gateway server Divertor/SOL 4 blocked all access except the IT-Based Laboratory Plasma-material interaction 3 InfraStructure (ITBL-IS) and Atomic Energy Grid Infrastructure (AEGIS) client certificate, and we Diagnostics 3 confirmed that its authentication mechanism was Heating system 2 working properly. The use of packet capture software Total 27 proved that packets were encrypted. The amount of communication data to display a discharge condition reference page was measured with packet capture software. Throughput was calculated from the window size and the measured Round Trip Time (RTT), and the turn around time was measured from the amount of communication data and throughput. When RTT was 290 milliseconds, the turn around time was 4.13 seconds for the remote experiment from IPP Garching. This gave the applicable response to the remote participants, and it provided mostly the same environment as the onsite researcher. Results are great advances towards the remote experiment in ITER. Fig.I.3.2-1 Remote experiment of JT-60 from IPP Garching. II. Theory, Simulation and Modeling 350 For the NEXT (Numerical Experiment of Tokamak) 300300 project in 2007, a new conservative gyrokinetic full-f 250 Vlasov code has been developed in order to realize a J long time simulation. Zonal field generation was 200200 + max + investigated in finite beta ion temperature gradient θ 150 driven turbulence by global Landau-fluid simulations. Jmin 100100 - - Also, nonlinear MHD simulations found the Alfven resonance effects on the evolution of magnetic islands 50 driven by an externally applied perturbation. The 00 0.7 0.75 0.8 0.85 0.9 0.95 distribution of eigenvalues of the resistive MHD 0.7 0.75 0.8 0.85 0.9 0.95 r/a equations in the complex plane has been re-investigated Fig. II.1.1-1 Two-dimensional profile of the perturbed current for smaller resistivity than the previous works, and a obtained for the rotation frequency Ω=0.02π. numerical matching scheme for linear MHD stability 0.9 analysis was proposed in a form offering tractable numerical implementation. A conjugate variable method Alfven resonance was formulated in order to apply the Hamilton-Lie 0.85 Jmin perturbation theory to a system of ordinary differential equations that does not have the Hamilton dynamic Jmax r/a q=2 structure. Jmin The development of the integrated modeling is 0.8 J promoting in order to study complex behaviors of max JT-60U plasmas and to predict the performance of future Alfven resonance burning plasmas. Several kinds of element codes have 0.75 been developed for the integrated modeling; MHD 0 500 1000 1500 2000 2500 stability code, integrated SOL/divertor code, t(τpa) fundamental SOL/divertor code of the particle model, Fig. II.1.1-2 Time evolution of the radial position of the heating and current drive analysis code, and integrated Alfven resonance and the maximum and minimum of the perturbed current. transport code TOPICS-IB. Achievements in 2007 with the use of these codes are described below in section 2. Alfven resonance moves to the magnetic neutral surface and causes the rapid growth of a driven magnetic island 1. Numerical Experiment of Tokamak (NEXT) (Fig.II.1.1-2). This is because the background flow is 1.1 Magnetohydrodynamic (MHD) Theory and damped by the appearance of the magnetic island, Simulation which operates torque on the plasma. These features are 1.1.1 Alfven Resonance Effects on an Externally Driven inconsistent with former theoretical assumptions, which Magnetic Island in Rotating Plasma enable using the asymptotic matching method to The Alfven resonance effects on the magnetic island estimate the force balance and the critical value of the evolution driven by an externally applied perturbation external perturbation, beyond which the driven were investigated. In a low viscosity regime, perturbed magnetic island grows rapidly [1.1-1]. current sheets form at the Alfven resonance surfaces (Fig.II.1.1-1), which differ from the radial position of 1.1.2 MHD Spectrum of Resistive Modes the magnetic neutral surface. Therefore, the sheets exist The resistive MHD spectrum is investigated in detail by outside the inner non-ideal layer, defined for the solving the eigenvalue problem of the reduced MHD non-rotating plasma. According to this perturbed current equations in cylindrical tokamak plasmas, in particular sheet profile, the total torque, which affects the plasma, for asymptotically smaller resistivity than the previous extends wider than the radial position of the Alfven works. In the presence of the resistivity, the eigenvalues resonance. It is found that the radial position of an of linear resistive MHD modes are not necessarily system with the Hamilton dynamic structure by purely real, but have an imaginary part. The eigenvalues doubling the unknown variables; hence the canonical are classified as the continuum spectrum of dumping Hamilton-Lie perturbation theory becomes applicable to modes on the negative real axis, the discrete spectrum of the system. The method was applied to two classical oscillatory and dumping modes on the complex plane, problems known in plasma physics to demonstrate the and the discrete spectrum of unstable modes on the effectiveness and to study the property of the method: positive real axis. For a wide range of the resistivity one is a non-linear oscillator that can explode; the other parameter (η=10-4 ~ 1.5×10-6), the shape and location of is the guiding center motion of a charged particle in a eigenvalue distribution for m/n=1/1 modes which have magnetic field [1.1-4]. the same resonant surface at q=1 is almost independent of the resistivity, although the density of the eigenvalues References increases. However, it is found that for a further small 1.1-1 Ishii, Y., et al., "Formation and long term evolution of resistivity regime, the eigenvalue distribution changes the externally driven magnetic island in rotating plasmas," to be published in Plasma and Fusion sensitively depending on the resistivity [1.1-2]. Research. 1.1-2 Matsumoto, T., et al., Proc. 6th General Scientific 1.1.3 Numerical Matching Scheme for Linear MHD Assembly of the Asia Plasma and Fusion Association Stability Analysis (India, 2007). A new matching scheme for linear MHD stability 1.1-3 Kagei, Y., et al., "Numerical matching scheme for analysis is proposed in a form offering tractable linear magnetohydrodynamic stability analysis," to be numerical implementation. This scheme divides the published in Plasma and Fusion Research. 1.1-4 Tokuda, S., et al., "Conjugate variable method in the plasma region into outer regions and inner layers, as in Hamilton-Lie perturbation theory -applications to the conventional matching method. However, the outer plasma physics-,” submitted to Plasma and Fusion regions do not contain any rational surface at their Research. terminal points; an inner layer contains a rational surface as an interior point. The Newcomb equation is 1.2 Plasma Turbulence Simulation therefore regular in the outer regions. The MHD 1.2.2 Development of Non-Dissipative Conservative equation employed in the layers is solved as an Finite Difference Scheme for Gyrokinetic Vlasov evolution equation in time, and the full implicit scheme Simulation is used to yield an inhomogeneous differential equation A new conservative gyrokinetic full-f Vlasov code is for space coordinates. The matching conditions are developed using a finite difference operator which derived from the condition that the radial component of conserves both the L1 and L2 norms. The growth of the solution in the layer is smoothly connected to those numerical oscillations is suppressed by conserving the in the outer regions at the terminal points. The proposed L2 norm, and the code is numerically stable and robust scheme was applied to the linear ideal MHD equation in in a long time simulation. In the slab ITG turbulence a cylindrical configuration, and was proved to be effective from the viewpoint of a numerical scheme [1.1-3]. 1.1.4 Conjugate Variable Method in the Hamilton-Lie Perturbation Theory -Applications to Plasma Physics- The conjugate variable method, which is an essential Fig. II.1.2-1 Comparisons of the time histories of the total, ingredient in the path-integral formalism of classical field, and kinetic energy in long time ITG turbulence statistical dynamics, was used in order to apply the simulations using (a) Vlasov and (b) PIC simulations with Hamilton-Lie perturbation theory to a system of almost the same computational costs. In (a), the total particle number is exactly conserved and the total energy conservation ordinary differential equations that does not have the is dramatically improved using a new conservative scheme. Hamilton dynamic structure. The method endows the simulation, the energy conservation and the entropy 2. Integrated Modeling balance relation are confirmed, and solutions are 2.1 MHD Stability − Effect of Equilibrium benchmarked against a conventional δf particle-in-cell Properties on the Structure of the Edge MHD (PIC) code. The results show that the exact particle Modes in Tokamaks − number conservation and the good energy conservation The ideal MHD stability code MARG2D has been in the conservative Vlasov simulation are advantageous extended to estimate a growth rate of the MHD mode for a long time micro-turbulence simulation (Fig. under the incompressible assumption by introducing the II.1.2-1). In the comparison, physical and numerical plasma inertia [2.1-1]. With this extension, MARG2D effects of the v// nonlinearity are clarified for the Vlasov realizes not only to identify the stability boundary of and PIC simulations [1.2-1, 2]. ideal MHD modes, but also to investigate physical properties of unstable MHD modes in detail. 1.2.1 Zonal Field Generation and Its Effects on Zonal By using this extended MARG2D, effects of the Flow and Turbulent Transport pressure profile and the current density profile inside the Global Landau-fluid simulations of ion temperature top of pedestal and that of the plasma shape on the gradient (ITG) driven turbulence have been performed expansion of the structure of the unstable edge MHD for finite beta tokamak plasmas, where the beta value is mode are investigated numerically [2.1-2]. The radial a ratio of plasma pressure to magnetic pressure. The structure of the edge MHD mode is expanded by ITG turbulence can drive zonal magnetic fields as well spreading the envelope of the edge ballooning mode due as zonal flows in finite beta cases. The Reynolds stress to increasing the pressure gradient inside the top of drives zonal flows and the geodesic transfer effect acts pedestal. Moreover, the increase of the current density as a sink for zonal flows usually. It is found that the induces the decrease in the toroidal mode number of the Reynolds stress and the geodesic transfer effect change most unstable mode, and this decrease also expands the their roles at low order rational surfaces where zonal structure of the unstable mode. magnetic fields are generated the most strongly. This is The mode structure is subject to expanding in not observed in electrostatic (zero beta) simulations. strongly shaped plasmas. This is because the pressure Effects of the zonal magnetic fields on the zonal flows gradient inside the top of pedestal can approach to the and the turbulent transport are limited in a small region ballooning mode stability boundary and the current around a low order rational surface at least in a low beta density increases enough to reduce the toroidal mode regime where the ITG mode is dominant. The zonal number of the most unstable mode. These increases of magnetic fields, however, may affect the zonal flows and the turbulent transport in a high beta regime Fig. II.2.1-1 (a) Profiles of plasma pressure because amplitude of the zonal magnetic fields (solid line) and parallel increases with the beta value [1.2-3]. current density (broken line) for three kinds of References equilibria. The current 1.2-1 Idomura, Y., et al., J. Comput. Phys. 226, 244 (2007). density inside the top of 1.2-2 Idomura, Y., et al., Commun. Nonlinear Sci. Numer. pedestal increases due Simul. 13, 227 (2007). to the steep pressure 1.2-3 Miyato, N., et al., Proc. 34th EPS Plasma Phys. Conf. gradient. (b) Radial (Poland, 2007), p4.043. structures of the most unstable MHD mode. Larger pressure gradient and current density inside the top of pedestal expand the structure of the unstable mode. the pressure gradient and the current density destabilize simulation code PARASOL has also been developed. the edge MHD mode and expand the mode structure. Simulation studies using those codes are progressed with the analysis of JT-60U experiments and the References divertor designing of JT-60SA. The X-point MARFE in 2.1-1 Aiba, N., et al., J. Plasma Fusion Res. 2, 010 (2007). the JT-60U experiment is simulated. It is found that the 2.1-2 Aiba, N., et al., to be published in J. Phys. Conf. Series deep penetration of chemically sputtered carbon at the (2008). dome for the detached phase causes the large radiation 2.2 SOL-Divertor peaking near the X-point as shown in Fig. II.2.2-1. The Divertor of tokamak reactors has four major functions, pumping capability of JT-60SA is evaluated through the heat removal, helium ash exhaust, impurity retention, simulation. A guideline to enhance the pumping and density control. Such divertor performance strongly efficiency is obtained in terms of the exhaust slot width depends on the various physics, i.e. plasma transport, and the strike point distance. Transient behavior of kinetic effects, atomic processes, and plasma-wall SOL/divertor plasmas after an ELM crash is interactions. In order to understand complicated divertor characterized by the PARASOL simulation; the physics and to predict divertor performance, JAEA has fast-time-scale heat transport is affected by collisions developed a series of divertor codes, onion-skin while the slow-time-scale behavior is affected by the recycling. modeling, SOLDOR, NEUT2D, IMPMC, PARASOL, 5-point divertor code coupled with a core transport code (TOPICS-IB), and Core-SOL-Divertor model (CSD). The benchmark test of SOLDOR/NEUT2D code with B2/EIRENE code was attempted. The simulation study of JT-60SA divertor was carried out with B2/EIRENE and the difference between single-null and double-null configurations was confirmed [2.2-1]. The CSD model was developed to take into account impurity radiation and momentum loss in the divertor region and the divertor detachment was investigated [2.2-2]. 2.2.1 Development of Integrated SOL/Divertor Code and Simulation Study To predict the heat and particle controllability in the divertor of tokamak reactors and to optimize the divertor design, comprehensive simulations by integrated modeling allowing for various physical Fig. II.2.2-1 Simulation results of (a) carbon radiation profile processes are indispensable. SOL/divertor codes have and (b) C+ ionization point distribution for X-point MARFE in been developed in Japan Atomic Energy Agency for the JT-60U. interpretation and the prediction on behavior of SOL/divertor plasmas, neutrals and impurities [2.2-3]. 2.2.2 Extension of IMPMC Code toward Time The code system consists of the 2D fluid code Evolution Simulation SOLDOR, the neutral Monte-Carlo (MC) code As a self-consistent modelling of divertor plasma and NEUT2D, and the impurity MC code IMPMC. Their impurity transport, the SONIC code package has been integration code “SONIC” is almost completed and developed. The key feature of this integrated code is to examined to simulate self-consistently the SOL/divertor incorporate the impurity MC code, IMPMC. The MC plasmas in JT-60U. In order to establish the physics approach is suitable for modelling of interactions modeling used in fluid simulations, the particle between impurities and walls, including kinetic effects, and the complicated dissociation process of and that to the far divertor plate become asymmetric. hydro-carbons. The MC modelling, however, has the The peak heat flux to the near plate is larger as disadvantage for long computational time, large MC compared to the far plate. The asymmetry in the peak noise, and assumption of steady state. The first and heat flux increases with the connection-length ratio. The second difficulties were solved by developing a new imbalance in the heat deposition, however, is small. The diffusion model and optimizing with a Message Passing radial transport loss of ELM flux creates the asymmetry Interface (MPI) on the massive parallel computer. The in the heat deposition, but the imbalance is still not large third subject is solved by extension of IMPMC code even for the large radial transport loss rate. The electron toward time evolution simulation. In time-dependent heat flux to the far plate brought by the SOL current is simulation with the MC code, a serious problem to one of the causes of a small imbalance in the heat increase number of test particles. The particle reduction deposition. Another cause is the asymmetric SOL flow method consisting of sorting the weights, pairing and and its convective heat flux, whose stagnation point Russian roulette has been proposed [2.2-4]. Sorting of stays for a long period near the ELM crash location. the weights is indispensable to suppress the MC noise. The divertor configuration in JT-60SA has been Reference 2.3-1 Takizuka T., et al., Contrib. Plasma Phys. 48, 207 optimized from a viewpoint of the neutral recycling (2008). with SOLDOR/NEUT2D. In the near future, it will be further optimized from a viewpoint of the impurity 2.4 Heating and Current Drive − Electron control with the SONIC code package coupled with the Cyclotron Current Drive in Magnetic Islands above extended IMPMC. of Neo-classical Tearing Mode − References Electron cyclotron current drive (ECCD) is the effective 2.2-1 Suzuki, Y., et al., Contrib. Plasma Phys. 48, 169 method for stabilization of neo-classical tearing modes (2008). (NTM). The driven current is evaluated conventionally 2.2-2 Hiwatari, R., et al., Contrib. Plasma Phys. 48, 174 by the bounce averaged Fokker-Planck equation (2008). (BAFP), where the magnetic field is assumed to be 2.2-3 Kawashima, H., et al., Plasma Phys. Control. Fusion 49, axi-symmetric. When the magnetic islands are formed S77 (2007). by NTMs, however, this assumption is incomplete and 2.2-4 Shimizu, K., et al., Contrib. Plasma Phys. 48, 270 (2008). the validity of ECCD analysis based on the BAFP equation becomes questionable. 2.3 ELM Transport − Effect of Radial Transport The ECCD in the magnetic island is studied Loss on the Asymmetry of ELM Heat Flux − numerically by a particle simulation [2.4-1]. Drift Large heat load on the divertor plate intermittently motion of electrons with Coulomb collisions and produced by ELMs is one of crucial issues for the tokamak fusion reactor research and development. The imbalance in the ELM heat loads on in-out divertor plates is also the problem. It has been reported that the ELM heat deposition to the outer plate is larger than that to the inner plate in JT-60U, while in JET and ASDEX Upgrade the inner-plate heat deposition becomes larger than the outer-plate heat deposition. To understand the physics background of such complex behaviors of the ELM heat flux, the dynamics of SOL-divertor plasmas after an ELM crash is studied with a one-dimensional Fig. II.2.4-1 EC power dependence of EC current drive particle simulation code, PARASOL [2.3-1]. efficiency with the magnetic island (solid circle) and without The ELM crash occurs off-centrally in the SOL island (open circle). The maximum of driven current density, region, and the ELM heat flux to the near divertor plate Jmax, is also shown by dashed line. velocity diffusion due to the EC waves is tracked by model for SOL and divertor plasmas. In the previous using a Monte-Carlo technique. An EC resonance region study, the TOPICS-IB successfully simulated transient is located around the O-point and localized along the behaviors of an H-mode plasma and clarified the toroidal direction. It is found that the driven current is mechanism of the collisionality dependence of the ELM well confined in the helical flux tube including the EC energy loss. The ELM energy loss, however, was less resonance region and the current channel looks like a than 10% of the pedestal energy. helical “Snake”. Figure II.2.4-1 shows the dependence The effect of the pressure gradient inside the of EC current drive efficiency I/P on the EC power P. pedestal top, P'inped, on the ELM energy loss is Plasma parameters are the followings; the magnetic examined [2.5-1]. Figure II.2.5-1(a) shows profiles of island of m/n = 2/1 is located around 0.5m in the the total pressure, P, at the ELM onset. The transport is 19 -3 minor-radius direction, where ne = 3×10 m , Te = reduced to the ion neoclassical level in the pedestal 10keV and Zeff = 1. The major radius is 3.5m and the EC region for case A, and the transport is additionally wave frequency is 125GHz (2nd harmonic resonance). reduced inside the pedestal top for other cases B-D. The The current drive efficiency with the magnetic island is pedestal top is located at ρ = 0.925 for all cases and larger than that without magnetic island and is enhanced P'inped becomes steeper in order of A, B, C, D. Even for by the increase of the EC power. The driven current the case A, P'inped is as the same as that observed in profile in the magnetic island becomes steep around the JT-60U. Profiles of the ELM enhanced diffusivity, χELM, O-point with the increase of P. These results are caused in the cases A-D are shown in Fig. II.2.5-1(b). The steep by the good confinement of EC resonant electrons pressure gradient broadens eigenfunctions of unstable inside the island like “Snake” and the nonlinear kinetic modes and the region of the ELM enhanced transport. effect due to the high EC power density. Figure II.2.5-1(c) shows the ELM energy loss, ΔWELM, normalized by the pedestal energy, Wped, as a function of Reference P'inped/P'ped where P'ped is the pedestal pressure gradient. 2.4-1 Hamamatsu K., et al., Plasma Phys. Control. Fusion 49, In the case A, the ELM energy loss is less than 10% of 1955 (2007). the pedestal energy and is comparable with those in JT-60U. The steep pressure gradient inside the pedestal 2.5 Integrated Simulation top enhances the ELM energy loss. The density collapse, Integrated simulation models have been developed on which is not considered here, enhances the values of the basis of the research in JT-60U experiments and first-principle simulations in order to clarify complex features of reactor-relevant plasmas. The integrated model of edge-pedestal, SOL and divertor clarified that the steep pressure gradient inside the H-mode pedestal top enhances the ELM energy loss [2.5-1]. A new one-dimensional core transport code, which can describe the radial electric field and plasma rotations, has been developed [2.5-2]. Success in these analysis and development leads to the further effective study of complex plasmas and methods to control the integrated performance. 2.5.1 Integrated ELM Simulation with Edge MHD Stability and Transport of SOL-Divertor Plasmas The energy loss due to ELMs has been investigated by using an integrated simulation code TOPICS-IB based Fig. II.2.5-1 (a) Pressure profiles at ELM onset and (b) χ on a 1.5 dimensional core transport code with a stability ELM for cases A-D. (c) Dependence of ΔWELM/Wped on P'inped/P'ped. code for the peeling-ballooning modes and a transport Definition positions of P'inped and P'ped are shown in (a). ΔWELM/Wped by about 50% under the assumption 2.5-1 Hayashi, N., et al., Contrib. Plasma Phys. 48, 196 of the similar collapse to the temperature one. The (2008). 2.5-2 Honda, M., et al., J. Comput.Phys. 227, 2808 (2008). ELM energy loss in the simulation becomes larger than 15% of the pedestal energy, as is shown in the database of multi-machine experiments. 2.5.2 Dynamic Transport Simulation Code Including Plasma Rotation and Radial Electric Field A new one-dimensional multi-fluid transport code named TASK/TX has been developed [2.5-2]. This code is able to describe dynamic behavior of tokamak plasmas, especially the time-evolution of the radial electric field and the plasma rotations. A set of flux-surface averaged equations is solved self-consistently in the cylindrical coordinates; Maxwell’s equations, continuity equations, equations of motion, heat transport equations, momentum transfer equations for fast particles, and two-group neutral diffusion equations. The finite element method with a piecewise linear interpolation function is employed. The Streamline Upwind Petrov–Galerkin method is also incorporated for numerically robust calculation. Despite solving the very nonlinear equations, the code shows a good convergence performance. Modification of a density profile during neutral beam injection (NBI) is presented. We found the density peaking for the counter-NBI and the density broadening for the co-NBI. The balance between neoclassical and turbulent effects defines the status of the density profile. We have confirmed that the TASK/TX well reproduces the profiles observed in JFT-2M. In the presence of ion orbit losses, the code predicts the generation of the intrinsic (spontaneous) rotation in the counter direction with the inward radial electric field. The non-ambipolar loss breaks quasi-neutrality and the plasma instantaneously generates the inward radial current near the periphery to compensate the ion loss current, inducing the more negative radial electric field and the torque toward the counter direction. Other conventional transport codes assuming the quasi-neutrality cannot follow these processes. It is the very special characteristic of the TASK/TX code that there is no need to impose an explicit quasi-neutrality condition. References 3. Atomic and Molecular Data We have been producing, collecting and compiling cross-section data for atomic and molecular collisions and spectral data relevant to fusion research. The electron capture and the electron loss cross-sections of singly ionized tungsten by collision with CH4 and C2H6 have been measured at collision energies of 27.2 and 54.4 eV/amu. The state selective charge transfer cross-section data of B5+ and C6+ by collision with H* ( n = 2 ) in the collision energy range between 62 eV/amu and 6.2 keV/amu have been calculated with a molecular-bases close-coupling method. The cross-section data for 42 processes of collisions of He, He*, He-, He+, He2+ and 3He2+ with H, - + H , H2, He and He have been complied. The recommended cross-section data are expressed with analytic functions to facilitate practical use of the data. The compiled data are in preparation for the Web at the URL http://www-jt60.naka.jaea.go.jp/english/JEAMDL/. The charge transfer data published in 2007 have been collected, and the database for the chemical sputtering yield data of graphite materials with hydrogen isotope collisions have been established. III. Fusion Reactor Design Study design value for various density and temperature 1. Progress in Compact DEMO Reactor Study profiles. For example, suppose that the BS current The design study of fusion DEMO reactor based on a around an internal transport barrier is dominant. Then slim CS (center solenoid coil) concept has been fBS is strongly dependent on q-value around the ITB, continued and still now be in progressing way. The i.e., the total current driven inside the ITB. For this slim CS concept leads to low aspect ratio plasma. The sense, q-profile control can be key technology to lower aspect ratio plasma is considered to have higher maintaining fBS at a design value in fusion plasma plasma performance (high elongation plasma and high especially with high fBS. In connection to this, the beta) and then is considered to be preferable from interplay between q-profile and pressure profile viewpoint of economical aspect. The size of CS coil (including ITB structure) will be an essential issue (radius of 75cm) is decided from the minimum governing the controllability on fBS. A concern about magnetic flux requirement of 20 volt-second. the analysis is consistency between the obtained The SlimCS reactor produces a fusion output of q-profile and the given density/temperature profiles. 2.95GW with a major radius of 5.5m, aspect ratio (A) This is an open question to be resolved with further of 2.6, normalized beta (βN) of 4.3 and maximum understanding on plasma transport. The result that even toroidal field of 16.4T. 1.5 MeV NB deposits in the peripheral region suggests Among the design progress issues, two issues the necessity to reconsider the role and beam energy of concerning a plasma profile consistency and a tritium NBI. A possible idea for it is to use lower energy (e.g. breeding are given in the followings. ~0.5 MeV) NBI for driving plasma rotation as well as Since plasma parameters are determined by a peripheral beam current. systems code based on a point model, the parameters should be checked for correctness using a one-dimensional (1-D) code. For this purpose, an ACCOME code [1-1] was used to investigate the consistency between the assumed plasma profiles and key parameters such as Pfus, βN, n/nGW, HHy2 and fBS. In the 1-D analysis, we attempted to find a solution with q-profile other than strongly reversed shear (RS). This is because strong RS does not seem to be appropriate as the standard operation mode of SlimCS from the points of view of disruptivity and the controllability of q in the central region. Figure III.1-1 shows a solution for weakly RS [1-2]. Although the profile is not reasonably optimized because of a single ECRF beamline, the following information was obtained from the analysis: 1) Most of the design parameters by the point model are consistent with the 1-D calculation; 2) The location of NBCD is restricted to the peripheral region because of beam attenuation even for 1.5 MeV NB; Fig.III.1-1 Example of weakly RS with consistency between 3) Use of ECRF as the main CD tool requires a high pressure and current profiles CD power due to its lower current drive efficiency than NBCD. Another issue is a systematic analysis of tritium In addition, the calculation indicates that q-profile breeding ratio (TBR) based on a one-dimensional control in the central and peripheral regions is neutronics code THIDA with FENDL2.0 library. important to maintain the bootstrap fraction around the Findings of the analysis are: • The required TBR is satisfied when 6Li is enriched to 2. Numerical Study on Beta Limit in Low Aspect about 70% for water-cooled solid pebble breeding Ratio Tokamak blanket consisting of Li2TiO3 and Be; The critical beta of low aspect ratio tokamak for • When 6Li is enriched up to 40% or more, there is no toroidal mode number n=1 is analyzed by using difference about TBR between Li2TiO3 and Li2ZrO3; MARG2D code developed in JAEA, which solves • Sector-wide copper conducting shell with 1cm in 2-dimensional Newcomb equation as an eigen-value thickness, which is placed in between replaceable problem [2-1]. and permanent blanket for vertical stability and high The maximum stable beta value is considered to be beta access of plasma, can be replaced by 7cm-thick mainly limited by RWM mode that is induced by ideal reduced activated martensitic steel (F82H) with a kink ballooning mode, and we analyze the ideal kink slight decrease in TBR. This means that the torus ballooning mode as a first step. As for the equilibria, configuration of SlimCS can be simplified by a typical up-down symmetric configuration of SlimCS modification of the structure of permanent blanket (elongation : κ=2.0, triangularity : δ=0.36) is used, and made of F82H. the profiles of plasma current and pressure are obtained When the actual tritium production is larger than from the correlation function using the experimental expected, for example by 5%, surplus production of profile data of JT-60U plasma with ITB (Internal tritium amounts to about 25 g/day, which will be Transport Barrier) [2-2], where the minimum values of extracted from the fuel cycle system and stored in the safety factor are set to be greater than 3. Critical on-site fuel storage. Since the surplus production of normalized beta 2.7 is obtained in the case of no ideal tritium reaches 9 kg for one-year operation, an in-situ wall. To obtain stable βN=4.3 SlimCS plasma, the TBR control method will be required to avoid conformal ideal wall must be placed at the position of excessive production. Borated-water is promising for 1.2 times plasma minor radius (Fig.III.2-1). Profile the purpose in that water borated with 0.7 wt% of optimization is necessary to obtain higher critical H3BO3 reduces the local TBR by 0.07 as shown in normalized beta. Fig.III.1-2, corresponding to a reduction of 0.05 in the net TBR. This indicates that borated-water injection 5.5 into coolant is useful as in-situ TBR control. 5.0 N ! 4.5 Unstable 4.0 Critical 3.5 3.0 Stable 2.5 1.0 1.5 2.0 2.5 r /a wall Fig.III.2-1 Critical βN vs. minor radius of conformal ideal wall for n=1 mode. References Fig.III.1-2 Local TBR as function of the concentration of 2-1 Aiba N., et al., Comp. Phys. Commun. 175 269 (2006). th H3BO3 in coolant. 2-2 Kurita G., Nagashima K., et al 1998 15 Annual Meeting of Japan Society of Plasma Science and NuclearFusion Research. References 1-1 Tani K., et al., J. Comput. Phys. 98, 332(1990). 1-2 Tobita K., et al., Nucl. Fusion 47, 892(2007). Appendix A.1 Publication List (April 2007 – March 2008) A.1.1 List of JAERI/JAEA Report 1) Chida, T., Ida, M., Nakamura, H., et al., “Thermo-Structural Analysis of Backwall in IFMIF Lithium Target, 2,” JAEA-Technology 2007-048 (2007) (in Japanese). 2) Fujieda, H., Sugihara, M., Shimada, M., et al., “Studies on Representative Disruption Scenarios, Associated Electromagnetic and Heat Loads and Operation Window in ITER,” JAEA-Research 2007-052 (2007). 3) Hayashi, K., Nakagawa, T., Onose, S., et al., “Investigation and Design of the Dismantling Process of Irradiation Capsules Containing Tritium, 1; Conceptual Investigation and Basic Design,” JAEA-Technology 2008-010 (2008) (in Japanese). 4) Ichige, H., Honda, M., Sasaki, S., et al., “Development of Pellet Injector Using Screw Type Pellet Extruder; Improvement of Pellet Extruder for High Frequency and Long Duration, and its Test Results,” JAEA-Technology 2007-037 (2007) (in Japanese). 5) Ida, M., Nakamura, H., Chida, T., et al., “Review of JAEA Activities on the IFMIF Liquid Lithium Target in FY2006,” JAEA-Review 2008-008 (2008). 6) Ishii, K., Seki, M., Shinozaki, S., et al., “Power Injection Performance of the LH Antenna Tipped with Carbon Grills in JT-60U,” JAEA-Technology 2007-036 (2007) (in Japanese). 7) Ishikawa, M., Kondoh, T., Hayakawa, A., et al., “Detail Design of Microfission Chamber for Fusion Power Diagnostic on ITER,” JAEA-Technology 2007-062 (2007). 8) Iwai, Y., Hayashi, T., Kobayashi, K., et al., “Application of Tritium Behavior Simulation Code (TBEHAVIOR) to an Actual-Scale Tritium Handling Room,” JAEA-Research 2007-070 (2007). 9) Kajita, S., Hatae, T., Katsunuma, A., et al., “Design Study of the Optical System in the Port Plug for Edge Thomson Scattering Diagnostics for ITER,” JAEA-Technology 2007-064 (2008). 10) Kakudate, S., Shibanuma, K., “Singular Point Analysis During Rail Deployment into Vacuum Vessel for ITER Blanket Maintenance,” JAEA-Technology 2007-032 (2007). 11) Oshima, K., Okano, F., Honda, A., et al., “Development of Protection System for Power Supply Facilities in JT-60U P-NBI for Long Pulse Operation,” JAEA-Technology 2007-044 (2007) (in Japanese). 12) Sakata, S., Kiyono, K., Oshima, T., et al., “Progress of Data Processing System in JT-60 Utilizing the UNIX-Based Workstations,” JAEA-Technology 2007-039 (2007) (in Japanese). 13) Seimiya, M., and JT-60 Operation Team, “The Archives of Operational Achievements in JT-60,” JAEA-Technology 2007-049 (2007) (in Japanese). 14) Seki, Y., Tanigawa, H., Tsuru, D., et al., “Studies on Tritium Breeding Ratio for Solid Breeder Blanket Cooled by Pressurized Water Through Nuclear and Thermal Analyses,” JAEA-Technology 2007-067 (2008) (in Japanese). 15) Shimada, K., Omori, Y., Okano, J., et al., “Initial Design Study of a DC Power Supply System for JT-60SA,” JAEA-Technology 2008-031 (2008) (in Japanese). 16) Shimada, K., Terakado, T., Kurihara, K., “Design Study of an AC Power System for Additional Heating Facilities in JT-60SA,” JAEA-Technology 2008-022 (2008) (in Japanese). 17) Sueoka, M., Kawamata, Y., Kurihara, K., et al., “Development of the Plasma Movie Database System in JT-60,” JAEA-Technology 2008-021 (2008) (in Japanese). 18) Suzuki, S., Seki, M., Shinozaki, S., et al., “Improvement of the Protection Devices for JT-60U LHRF Antenna System,” JAEA-Technology 2007-055 (2007) (in Japanese). 19) Takato, N., Tobari, H., Inoue, T., et al., “Spatial Uniformity of Negative Ion Beam in Magnetically Filtered Hydrogen Negative Ion Source; Effect of the H- Ion Production and Transport Processes on the H- Ion Beam Intensity profile in the Cs-seeded negative ion source (Joint research),” JAEA-Research 2008-031 (2008) (in Japanese). 20) Terakado, M., Shimono, M., Sawahata, M., et al., “Development of the Power Modulation Technique in JT-60U ECH System,” JAEA-Technology 2007-053 (2007) (in Japanese). 21) Totsuka, T., “Development of the Java-Based Man-Machine Interfacing System for Remote Experiments on JT-60,” JAEA-Technology 2007-034 (2007) (in Japanese). 22) Yamaguchi, T., Kawano, Y., Kusama, Y., “Sensitivity Study of the ITER Poloidal Polarimeter,” JAEA-Research 2007-078 (2008). 23) Yamauchi, M., Hori, J., Sato, S., et al., “ACT-XN; Revised Version of an Activation Calculation Code for Fusion Reactor Analysis; Supplement of the Function for the Sequential Reaction Activation by Charged Particles,” JAEA-Data/Code 2007-016 (2007) (in Japanese). 24) Yokokura, K., Moriyama, S., Hasegawa, K., et al., “Development of Power Measuring Device of Transmission Type with Dielectric for High Power Millimeter Wave,” JAEA-Technology 2007-045 (2007) (in Japanese). A.1.2 List of Papers Published in Journals 1) Aiba, N., Tokuda, S., Fujita, T., et al., “Numerical Method for the Stability Analysis of Ideal MHD Modes with a Wide Range of Toroidal Mode Numbers in Tokamaks,” Plasma Fusion Res., (Internet) 2, 010 (2007). 2) Aiba, N., Tokuda, S., Takizuka, T., et al., “Effects of "Sharpness" of the Plasma Cross-Section on the MHD Stability of Tokamak Edge Plasmas,” Nucl. Fusion, 47, 297 (2007). 3) Araghy, H.P., Peterson, B.J., (Konoshima, S.), et al., “Spatial Variation of the Foil Parameters from in Situ Calibration of the JT-60U Imaging Bolometer Foil,” Plasma Fusion Res., (Internet) 2, S1116 (2007). 4) Asakura, N., ITPA SOL and Divertor Topical Group, “Understanding the SOL Flow in L-Mode Plasma on Divertor Tokamaks, and its Influence on the Plasma Transport,” J. Nucl. Mater., 363-365, 41 (2007). 5) Ashikawa, N., Kizu, K., (Miya, N.), et al., “Comparison of Boronized Wall in LHD and JT-60U,” J. Nucl. Mater., 363-365, 1352 (2007). 6) Batistoni, P., Angelone, M., (Ochiai, K.), et al., “Neutronics Experiment on a Helium Cooled pebble bed (HCPB) Breeder Blanket Mock-Up,” Fusion Eng. Des., 82, 2095 (2007). 7) Briguglio, S., Fogaccia, G., (Shinohara, K.), et al., “Particle Simulation of Bursting Alfvén Modes in JT-60U,” Physics of Plasmas, 14, 055904 (2007). 8) Callen, J.D., Anderson, J.K., (Fujita, T.), et al., “Experimental tests of paleoclassical transport,” Nucl. Fusion, 47, 1449 (2007). 9) Chankin, A.V., Coster, D.P., Asakura, N., et al., “Discrepancy Between Modelled and Measured Radial Electric Fields in the Scrape-Off Layer of Divertor Tokamaks; A Challenge for 2D Fluid Codes?,” Nucl. Fusion, 47, 479 (2007). 10) de Vries, P.C., Salmi, A., (Oyama, N.), et al., “Effect of Toroidal Field Ripple on Plasma Rotation in JET,” Nucl. Fusion, 48, 035007 (2008). 11) Donné, A.J.H., Costley, A.E., (Kawano, Y.), et al., “Progress in the ITER Physics Basis, 7; Diagnostics,” Nucl. Fusion, 47, S337 (2007). 12) Doyle, E.J., Houlberg, W.A., Kamada, Y., et al., “Progress in the ITER Physics Basis, 2; Plasma Confinement and Transport,” Nucl. Fusion, 47, S18 (2007). 13) Fasoli, A., Gormezano, C., (Shinohara, K.), et al., “Progress in the ITER Physics Basis, 5; Physics of Energetic Ions,” Nucl. Fusion, 47, S264 (2007). 14) Fujimoto, K., Nakano, T., Kubo, H., et al., “Modification of Tomography Technique for Two-Dimensional Spectroscopic Measurement in JT-60U Divertor Plasmas,” Plasma Fusion Res., (Internet) 2, S1121 (2007). 15) Fujino, I., Hatayama, A., (Inoue, T.), et al., “Analysis of Electron Energy Distribution of an Arc-Discharge H- Ion Source with Monte Carlo Simulation,” Rev. Sci. Instrum., 79, 02A510 (2008). 16) Fujisawa, A., Ido, T., Hoshino, K., et al., “Experimental Progress on Zonal Flow Physics in Toroidal Plasmas,” Nucl. Fusion, 47, S718 (2007). 17) Fujita, T., Tamai, H., Matsukawa, M., et al., “Design Optimization for Plasma Performance and Assessment of Operation Regimes in JT-60SA,” Nucl. Fusion, 47, 1512 (2007). 18) Gormezano, C., Sips, A.C.C., (Ide, S.), et al., “Progress in the ITER Physics Basis, 6; Steady State Operation,” Nucl. Fusion, 47, S285 (2007). 19) Gribov, Y., Humphreys, D.A., (Ozeki, T.), et al., “Progress in the ITER Physics Basis, 8; Plasma Operation and Control,” Nucl. Fusion, 47, S385 (2007). 20) Hamada, K., Nakajima, H., Kawano, K., et al., “Demonstration of Full Scale JJ1 and 316LN Fabrication for ITER TF Coil Structure,” Fusion Eng. Des., 82, 1481 (2007). 21) Hamamatsu, K., Takizuka, T., Hayashi, N., et al., “Numerical Simulation of Electron Cyclotron Current Drive in Magnetic Islands of Neo-Classical Tearing Mode,” Plasma Phys. Control. Fusion, 49, 1955 (2007). 22) Hanada, M., Ikeda, Y., Kamada, M., et al., “Correlation Between Voltage Holding Capability and Light Emission in a 500 keV Electrostatic Accelerator Utilized for Fusion Application,” IEEE Trans. Dielectr. Electr. Insulat., 143, 572 (2007). 23) Hanada, M., Inoue, T., Kashiwagi, M., et al., “R&D Progress at JAEA Towards Production of High Power and Large-Area Negative Ion Beams for ITER,” Nucl. Fusion, 47, 1142 (2007). 24) Hatae, T., Howard, J., Hirano, Y., et al., “Development of polarization interferometer based on Fourier transform spectroscopy for Thomson scattering diagnostics,” Plasma Fusion Res., (Internet) 2, S1026 (2007). 25) Hayashi, N., Takizuka, T., Aiba, N., et al., “Integrated ELM Simulation with Edge MHD Stability and Transport of SOL-Divertor Plasmas,” Contrib. Plasma Phys., 48, 196 (2008). 26) Hayashi, N., Takizuka, T., Hosokawa, M., et al., “Modeling of Dynamic Response of SOL-Divertor Plasmas to an ELM Crash,” J. Nucl. Mater., 363-365, 1044 (2007). 27) Hayashi, N., Takizuka, T., Ozeki, T., et al., “Integrated Simulation of ELM Energy Loss Determined by Pedestal MHD and SOL Transport,” Nucl. Fusion, 47, 682 (2007). 28) Hayashi, T., Kasada, R., Tobita, K., et al., “Impact of N-Isotope Composition Control of Ferritic Steel on Classification of Radioactive Materials from Fusion Reactor,” Fusion Eng. Des., 82, 2850 (2007). 29) Hayashi, T., Sakurai, S., Masaki, K., et al., “Conceptual Design of Divertor Cassette Handling by Remote Handling System of JT-60SA,” Journal of Power and Energy Systems (Internet), 2, 522 (2008). 30) Hayashi, T., Sugiyama, K., Krieger, K., et al., “Deuterium Depth Profiling in JT-60U Tiles Using the D(3He, p)4He Resonant Nuclear Reaction,” J. Nucl. Mater., 363-365, 904 (2007). 31) Hayashi, T., Isobe, K., Kobayashi, K., et al., “Recent Activities on Tritium Technologies for ITER and Fusion Reactors at JAEA,” Fusion Sci. Tech., 52, 651 (2007). 32) Hayashi, T., Nakamura, H., Isobe, K., et al., “Tritium Behavior on the Water-Metal Boundary for the Permeation Into Cooling Water Through Metal Piping,” Fusion Sci. Tech., 52, 687 (2007). 33) Hayashi, T., Suzuki, T., Shu, W., et al., “Isotope Effect of Hydrogen Rapidly Supplied from the Metal Storage Bed,” Fusion Sci. Tech., 52, 706 (2007). 34) Hender, T.C., Wesley, J.C., (Isayama, A.), et al., “Progress in the ITER Physics Basis, 3; MHD Stability, Operational Limits and Disruptions,” Nucl. Fusion, 47, S128 (2007). 35) Hirohata, Y., Tanabe, T., (Miya, N.), et al., “Hydrogen Isotopes Retention in JT-60U,” J. Nucl. Mater., 363-365, 854 (2007). 36) Hiroki, S., Tanzawa, S., Arai, T., et al., “Development of Water Leak Detection Method in Fusion Reactors Using Water-Soluble Gas,” Fusion Eng. Des., 83, 72 (2008). 37) Hirose, T., Tanigawa, H., Enoeda, M., et al., “Effects of Tube Drawing on Structural Material for ITER Test Blanket Module,” Fusion Sci. Tech., 52, 839 (2007). 38) Honda, M., Fukuyama, A.,㻃 “Dynamic Transport Simulation Code Including Plasma Rotation and Radial Electric Field,” J. Comput. Phys., 227, 2808 (2008). 39) Hoshino, K., Suzuki, T., Isayama, A., et al., “Electron Cyclotron Heating Applied to the JT-60U Tokamak,” Fusion Sci. Tech., 53, 114 (2008). 40) Hoshino, T., Yasumoto, M., Tsuchiya, K., et al., “Non-Stoichiometory and Vaporization Characteristic of Li2.1TiO3.05 in Hydrogen Atmosphere,” Fusion Eng. Des., 82, 2269 (2007). 41) Hosogane, N., JT-60SA Design Team and Japan-Europe Satellite Tokamak Working Group, “Superconducting Tokamak JT-60SA Project for ITER and DEMO Researches,” Fusion Sci. Tech., 52, 375 (2007). 42) Ichimura, M., Higaki, H., (Moriyama, S.), et al., “Observation of Spontaneously Excited Waves in the Ion Cyclotron Frequency Range on JT-60U,” Nucl. Fusion, 48, 035012 (2008). 43) Ida, M., Nakamura, H., Sugimoto, M., “Analytical Estimation of Accessibility to the Activated Lithium Loop in IFMIF,” J. Nucl. Mater., 367-370, 1557 (2007). 44) Ida, M., Nakamura, H., Sugimoto, M., “Estimation and Control of Beryllium-7 Behavior in Liquid Lithium Loop of IFMIF,” Fusion Eng. Des., 82, 2490 (2007). 45) Ide, S., Takenaga, H., Isayama, A., et al., “Studies on Impact of Electron Cyclotron Wave Injection on the Internal Transport Barriers in JT-60U Weak Shear Plasmas,” Nucl. Fusion, 47, 1499 (2007). 46) Idomura, Y., Ida, M., Tokuda, S., “Conservative Gyrokinetic Vlasov Simulation,” Commun. Nonlinear. Sci. Numer. Simulat., 13, 227 (2008). 47) Idomura, Y., Ida, M., Tokuda, S., et al., “New conservative gyrokinetic full-f Vlasov code and its comparison to gyrokinetic δf particle-in-cell code,” J. Comput. Phys., 226, 244 (2007). 48) Ikeda, Y., Akino, N., Ebisawa, N., et al., “Technical design of NBI system for JT-60SA,” Fusion Eng. Des., 82, 791 (2007). 49) Inoue, T., Hanada, M., Kashiwagi, M., et al., “Development of Beam Source and Bushing for ITER NB System,” Fusion Eng. Des., 82, 813 (2007). 50) Inoue, T., Tobari, H., Takado, N., et al., “Negative Ion Production in Cesium Seeded High Electron Temperature Plasmas,” Rev. Sci. Instrum., 79, 02C112 (2008). 51) Ioki, K., Elio, F., Barabash, V., et al., “Six-Party Qualification Program of FW Fabrication Methods for ITER Blanket Module Procurement,” Fusion Eng. Des., 82, 1774 (2007). 52) Isayama, A., Oyama, N., Urano, H., et al., “Stabilization of Neoclassical Tearing Modes by Electron Cyclotron Current Drive in JT-60U,” Nucl. Fusion, 47, 773 (2007). 53) Ishii, Y., Azumi, M., Smolyakov, A.I., “Nonliear Evolution and Deformation of Driven Magnetic Islands in Rotating Plasmas,” Nucl. Fusion, 47, 1024 (2007). 54) Ishikawa, M., Nishitani, T., Kusama, Y., et al., “Neutron Emission Profile Measurement and Fast Charge Exchange Neutral Particle Flux Measurement for Transport Analysis of Energetic Ions in JT-60U,” Plasma Fusion Res., (Internet) 2, 019 (2007). 55) Ishikawa, M., Takechi, M., Shinohara, K., et al., “Confinement Degradation and Transport of Energetic Ions due to Alfvén Eigenmodes in JT-60U Weak Shear Plasmas,” Nucl. Fusion, 47, 849 (2007). 56) Ito, T., Hayashi, T., Isobe, K., et al., “Self-Decomposition Behavior of High Concentration Tritiated Water,” Fusion Sci. Tech., 52, 701 (2007). 57) Jolliet, S., Bottino, A., (Idomura, Y.), et al., “A Global Collisionless PIC Code in Magnetic Coordinates,” Comput. Phys. Commun., 177, 409 (2007). 58) Kajita, S., Ono, N., Takamura, S., et al., “Plasma-Assisted Laser Ablation of Tungsten; Reduction in Ablation Power Threshold due to Bursting of Holes/Bubbles,” Applied Physics Letters, 91, 261501 (2007). 59) Kamiya, K., Asakura, N., Boedo, J.A., et al., “Edge Localized Modes; Recent Experimental Findings and Related Issues,” Plasma Phys. Control. Fusion, 49, s43 (2007). 60) Kanemura, T., Kondo, H., (Ida, M.), et al., “Investigation of Free-Surface Fluctuations of Liquid Lithium Flow for IFMIF Lithium Target by Using an Electro-Contact Probe,” Fusion Eng. Des., 82, 2550 (2007). 61) Kawashima, H., Shimizu, K., Takizuka, T., “Development of Integrated SOL/Divertor Code and Simulation Study of the JT-60U/JT-60SA Tokamaks,” Plasma Phys. Control. Fusion, 49, s77 (2007). 62) Kawashima, H., Shimizu, K., Takizuka, T., et al., “Simulation of Divertor Pumping in JT-60U with SOLDOR/NEUT2D Code,” J. Nucl. Mater., 363-365, 786 (2007). 63) Kizu, K., Tsuchiya, K., Ando, T., et al., “Conceptual Design of Magnet System for JT-60 Super Advanced (JT-60SA),” IEEE Trans. Appl. Superconduct., 17, 1348 (2007). 64) Kizu, K., Tsuchiya, K., Shimada, K., et al., “Evaluation of Bending Strain Dependence of Critical Current of Nb3Al Conductor for Coils with React-And-Wind Method,” Fusion Eng. Des., 82, 1493 (2007). 65) Kobayashi, K., Hayashi, T., Nakamura, H., et al., “Study for the Behavior of Tritiated Water Vapor on Organic Materials,” Fusion Sci. Tech., 52, 696 (2007). 66) Kobayashi, K., Isobe, K., Iwai, Y., et al., “Studies on the Behavior of Tritium in Components and Structure Materials of Tritium Confinement and Detritiation Systems of ITER,” Nucl. Fusion, 47, 1645 (2007). 67) Kobayashi, K., Miura, H., Hayashi, T., et al., “Oxidation Performance Test of Detritiation System Under Existence of SF6,“ Fusion Sci. Tech., 52, 711 (2007). 68) Kobayashi, S., and JT-60 Team, “Development of Real-Time Measurement System of Charge Exchange Recombination Spectroscopy and its Application to Feedback Control of Ion Temperature Gradient in JT-60U,” Plasma Fusion Res., (Internet) 2, S1049 (2007). 69) Kobayashi, T., Moriyama, S., Seki, M., et al., “Achievement of 1.5 MW, 1 s Oscillation by the JT-60U Gyrotron,” Plasma Fusion Res., (Internet) 3, 014 (2008). 70) Koizumi, N., Isono, T., Hamada, K., et al., “Development of Large Current Superconductors Using High Performance Nb3Sn Strand for ITER,” Physica C, 463-465, 1319 (2007). 71) Kojima, A., Kamiya, K., Iguchi, H., et al., “Numerical simulation of a high-brightness lithium ion gun for a Zeeman polarimetry on JT-60U,” Plasma Fusion Res., (Internet) 2, S1104 (2007). 72) Kondo, H., Kanemura, T., (Ida, M.), et al., “Measurement of Free Surface of Liquid Metal Lithium Jet for IFMIF Target,” Fusion Eng. Des., 82, 2483 (2007). 73) Kondo, K., Murata, I., Ochiai, K., et al., “Measurement and Analysis of Neutron-Induced Alpha Particle Emission Double-Differential Cross Section of Carbon at 14.2 MeV,” J. Nucl. Sci. Technol., 45, 103 (2008). 74) Kondo, K., Murata, I., Ochiai, K., et al., “Verification of Nuclear Data for DT Neutron Induced Charged-Particle Emission Reaction of Light Nuclei,” Fusion Eng. Des., 82, 2786 (2007). 75) Kubo, H., Sasaki, A., Moribayashi, K., et al., “Study of Highly Ionized Xe Spectra with 3s-3p and 3p-3d Transitions in JT-60U Reversed Shear Plasmas,” J. Nucl. Mater., 363-365, 1441 (2007). 76) Kubota, N., Kondo, K., Ochiai, K., et al., “Neutron Elastic Recoil Detection for Hydrogen Isotope Analysis in Fusion Materials,” J. Nucl. Mater., 367-370, 1596 (2007). 77) Li, Z., Tanaka, T., (Sato, S.), et al., “Spectral Effects of Activation for Liquid Blanket Relevant Materials Induced by D-T Neutron Irradiation,” Fusion Sci. Tech., 52, 817 (2007). 78) Liu, Y., Tamura, N., (Konoshima, S.), et al., “Application of Tomographic Imaging to Multi-Pixel Bolometric Measurements,” Plasma Fusion Res., (Internet) 2, S1124 (2007). 79) Loarte, A., Lipschultz, B., (Asakura, N.), et al., “Progress in the ITER Physics Basis, 4; Power and Particle Control,” Nucl. Fusion, 47, S203 (2007). 80) Maggi, C.F., Groebner, R.J., Oyama, N., et al., “Characteristics of the H-mode Pedestal in Improved Confinement Scenarios in ASDEX Upgrade, DIII-D, JET and JT-60U,” Nucl. Fusion, 47, 535 (2007). 81) Masaki, K., Tanabe, T., Hirohata, Y., et al., “Hydrogen Retention and Carbon Deposition in Plasma Facing Components and the Shadowed Area of JT-60U,” Nucl. Fusion, 47, 1577 (2007). 82) Matsumoto, T., Li, J., Kishimoto, Y., “Characteristics of ETG-Driven Turbulence Dominated by Zonal Flows,” Nucl. Fusion, 47, 880 (2007). 83) Matsushita, D., Takado, N., (Inoue, T.), et al., “Numerical Analysis of H- Ion Transport Processes in Cs-Seeded Negative Ion Sources,” Rev. Sci. Instrum., 79, 02A527 (2008). 84) Miki, K., Kishimoto, Y., Miyato, N., et al., “Intermittent Transport Associated with the Geodesic Acoustic Mode Near the Critical Gradient Regime,” Physical Review Letters, 99, 145003 (2007). 85) Mishima, Y., Yoshida, N., (Tsuchiya, K.), et al., “Recent Results on Beryllium and Beryllides in Japan,” J. Nucl. Mater., 367-370, 1382 (2007). 86) Miyato, N., Kishimoto, Y., Li, J.Q., “Turbulence Suppression in the Neighbourhood of a Minimum-q Surface due to Zonal Flow Modification in Reversed Shear Tokamaks,” Nucl. Fusion, 47, 929 (2007). 87) Morimoto, M., Ioki, K., Terasawa, A., et al., “Design Progress of the ITER In-Wall Shielding,” Fusion Sci. 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Fusion, 49, 1611 (2007). 93) Nakajima, H., Hamada, K., Okuno, K., et al., “Development of Optimum Manufacturing Technologies of Radial Plates for the ITER Toroidal Field Coils,” Fusion Eng. Des., 82, 1473 (2007). 94) Nakamichi, M., Kulsartov, T.V., Hayashi, K., et al., “In-pile Tritium Permeation Through F82H Steel with and Without a Ceramic Coating of Cr2O3-SiO2 Including CrPO4,” Fusion Eng. Des., 82, 2246 (2007). 95) Nakamura, H., Kobayashi, K., Yamanishi, T., et al., “Tritium Release Behavior from Steels Irradiated by High Energy Protons,” Fusion Sci. Tech., 52, 1012 (2007). 96) Nakamura, H., Ida, M., Chida, T., et al., “Design of a Lip Seal-Replaceable Backwall for IFMIF Liquid Lithium Target,” Fusion Eng. Des., 82, 2671 (2007). 97) Nakamura, T., Sakagami, H., (Koga, J.K.), et al., “High Energy Electron Generation by Laser-Cone Interaction,” Plasma Fusion Res., (Internet) 2, 018 (2007). 98) Nakamura, T., Sakagami, H., (Koga, J.K.), et al., “Optimization of Cone Target Geometry for Fast Ignition,” Physics of Plasmas, 14, 103105 (2007). 99) Nakano, T., and JT-60 Team, “Particle Balance Under Global Wall Saturation in Long-Pulse Discharges of JT-60U,” J. Nucl. Mater., 363-365, 1315 (2007). 100) Nakano, T., Kubo, H., Asakura, N., et al., “Volume Recombination of C4+ in Detached Divertor Plasmas of JT-60U,” Nucl. Fusion, 47, 1458 (2007). 101) Nishimura, A., Nishijima, S., (Nishitani, T.), et al., “Change in Properties of Superconducting Magnet Materials by Fusion Neutron Irradiation,” Fusion Eng. Des., 82, 1555 (2007). 102) Nishitani, T., Enoeda, M., Akiba, M., et al., “Recent Progress in Solid Breeder Blanket Development at JAEA,” Fusion Sci. Tech., 52, 971 (2007). 103) Nishitani, T., Sato, S., Ochiai, K., et al., “Progress in Neutronics Studies for the Water Cooled Pebble Bed Blanket,” Fusion Sci. Tech., 52, 791 (2007). 104) Nishitani, T., Yamauchi, M., Izumi, M., et al., “Engineering Design of the ITER Invessel Neutron Monitor Using Micro-Fission Chambers,” Fusion Eng. Des., 82, 1192 (2007). 105) Nunoya, Y., Isono, T., Koizumi, N., et al., “Development of Strain-Applying Apparatus for Evaluation of ITER Nb3Sn Strand,” IEEE Trans. Appl. Superconduct., 17, 2588 (2007). 106) Ochiai, K., Sato, S., Wada, M., et al., “Thin Slit Streaming Experiment for ITER by Using D-T Neutron Source,” Fusion Eng. Des., 82, 2794 (2007). 107) Ogawa, H., Sugie, T., Kasai, S., et al., “Development of Impurity Influx Monitor (Divertor) for ITER Plasma,” Fusion Res., (Internet) 2, S1054 (2007). 108) Ogawa, M., Iio, S., (Inoue, T.), et al., “Magnetic Fusion Energy Studies in Japan,” Nucl. Instrum. Methods. Phys. Res., A, 577, 30 (2007). 109) Ohira, S., Hayashi, T., Shu, W., et al., “Radiochemical Characteristics of Tritium to be Considered in Fusion Reactor Facility Design,” J. Radioanal. Nucl. Chem., 272, 575 (2007). 110) Oka, K., Onuki, S., Yamashita, S., et al., “Structure of Nano-Size Oxides in ODS Steels and its Stability Under Electron Irradiation,” Materials Transactions, 48, 2563 (2007). 111) Okuno, K., Nakajima, H., Sugimoto, M., et al., Technology Development for the Construction of the ITER Superconducting Magnet System,” Nucl. Fusion, 47, 456 (2007). 112) Onozuka, M., Shimizu, K., (Nakajima, H.), et al., “Basic Analysis of Weldability and Machinability of Structural Materials for ITER Toroidal Field Coils,” Fusion Eng. Des., 82, 1431 (2007). 113) Oshima, T., Kiyono, K., Sakata, S., et al., “Improvement of Data Processing System for Advanced Diagnostics in JT-60U,” Fusion Eng. Des., 82, 1210 (2007). 114) Oya, Y., Hirohata, Y., (Masaki, K.), et al., “Hydrogen Isotope Retentions and Erosion/Deposition Profiles in the First Wall of JT-60U,” Fusion Sci. Tech., 52, 554 (2007). 115) Oyama, N., and JT-60 Team, “Improved Performance in Long-Pulse ELMy H-mode Plasmas with Internal Transport Barrier in JT-60U,” Nucl. Fusion, 47, 689 (2007). 116) Ozeki, T., and JT-60 Team, “High-Beta Steady-State Research with Integrated Modeling in the JT-60 Upgrade,“ Physics of Plasmas, 14, 056114 (2007). 117) Peterson, B.J., LHD Team and JT-60U Team, “Research and Development of Imaging Bolometers,” Plasma Fusion Res., (Internet) 2, S1018 (2007). 118) Peterson, B.J., and JT-60 Team, “Observation of Divertor and Core Radiation in JT-60U by Means of Bolometric Imaging,” J. Nucl. Mater., 363-365, 412 (2007). 119) Prater, R., La Haye, R.J., (Isayama, A.), et al., “Stabilization and Prevention of the 2/1 Neoclassical Tearing Mode for Improved Performance in DIII-D,” Nucl. Fusion, 47, 371 (2007). 120) Rewoldt, G., Lin, Z., Idomura, Y., “Linear Comparison of Gyrokinetic Codes with Trapped Electrons,” Comput. Phys. Commun., 177, 775 (2007). 121) Saibene, G., Oyama, N., Lönnroth, J., et al., “The H-Mode Pedestal, ELMs and TF Ripple Effects in JT-60U/JET Dimensionless Identity Experiments,” Nucl. Fusion, 47, 969 (2007). 122) Sakamoto, K., “Progress of High-Power-Gyrotron Development for Fusion Research,” Fusion Sci. Tech., 52, 145 (2007). 123) Sakamoto, K., Kasugai, A., Takahashi, K., et al., “Achievement of Robust High-Efficiency 1 MW Oscillation in the Hard-Self-Excitation Region by a 170 GHz Continuous-Wave Gyrotron,” Nature Physics, 3, 411 (2007). 124) Sakamoto, Y., and JT-60 Team, “Controllability of Large Bootstrap Current Fraction Plasmas in JT-60U,” Nucl. Fusion, 47, 1506 (2007). 125) Sakamoto, Y., and JT-60 Team, “Controllability of Large Bootstrap Current Fraction Plasmas in JT-60U,” Nucl. Fusion, 47, 1506 (2007). 126) Sasao, M., Nishitani, T., Krasilnikov, A., et al., “Fusion Product Diagnostics,” Fusion Sci. Tech., 53, 604 (2008). 127) Sato, K., Omori, J., Ebisawa, K., et al., “Development of ITER Diagnostic Upper Port Plug,” Plasma Fusion Res., (Internet) 2, S1088 (2007). 128) Sato, M., Isayama, A., “Evaluation of Extended Trubnikov Emissivity to the Oblique Propagation and Application to Electron Temperature Measurement in a Reactor-Grade Tokamak,” Fusion Sci. Tech., 52, 169 (2007). 129) Sato, M., Isayama, A., “Effects of Relativistic and Absorption on ECE Spectra in High Temperature Tokamak Plasma,” Plasma Fusion Res., (Internet) 2, S1029 (2007). 130) Sato, S., Ochiai, K., Verzilov, Y., et al., “Measurement of Tritium Production Rate in Water Cooled Pebble Bed Multi-Layered Blanket Mockup by DT Neutron Irradiation Experiment,” Nucl. Fusion, 47, 517 (2007). 131) Sato, S., Verzilov, Y., Ochiai, K., et al., “Neutronics Experimental Study on Tritium Production in Solid Breeder Blanket Mockup with Neutron Reflector,” J. Nucl. Sci. Technol., 44, 657 (2007). 132) Seki, M., “ITER Activities and Fusion Technology,” Nucl. Fusion, 47, S489 (2007). 133) Shibama, Y., Arai, T., Miyo, Y., et al., “Structural Design of Ferritic Steel Tiles for Ripple Reduction of Toroidal Magnetic Field in JT-60U,” Fusion Eng. Des., 82, 2462 (2007). 134) Shimada, K., Ito, J., Matsukawa, M., “A Control Method of Matrix Converter for Plasma Control Coil power Supply,” Fusion Eng. Des., 82, 1513 (2007). 135) Shimada, M., Campbell, D.J., Mukhovatov, V., et al., “Progress in the ITER Physics Basis, 1; Overview and Summary,” Nucl. Fusion, 47, S1 (2007). 136) Shimizu, K., Takizuka, T., Kawashima, H., “A New Fast Velocity-Diffusion Modelling for Impurity Transport in Integrated Edge Plasma Simulation,” J. Nucl. Mater., 363-365, 426 (2007). 137) Shimizu, K., Takizuka, T., Kawashima, H., “Extension of IMPMC Code Toward Time Evolution Simulation,” Contrib. Plasma Phys., 48, 270 (2008). 138) Shimomura, K., Takenaga, H., Tsutsui, H., et al., “Burn Control Simulation Experiments in JT-60U,” Fusion Eng. Des., 82, 953 (2007). 139) Shinohara, K., Isobe, M., Darrow, D.S., et al., “Escaping Ion Measurement with High Time Resolution During Bursting Modes Induced by Neutral Beam Injection on CHS,” Plasma Fusion Res., (Internet) 2, 042 (2007). 140) Shinohara, K., and JT-60 Team, “Ferritic Insertion for Reduction of Toroidal Magnetic Field Ripple on JT-60U,” Nucl. Fusion, 47, 997 (2007). 141) Sueoka, M., Kawamata, Y., Kurihara, K., et al., “The Plasma Movie Database System for JT-60,” Fusion Eng. Des., 82, 1008 (2007). 142) Sugiyama, K., Hayashi, T., (Miya, N.), et al, “Ion Beam Analysis of H and D Retention in the Near Surface Layers of JT-60U Plasma Facing Wall Tiles,” J. Nucl. Mater., 363-365, 949 (2007). 143) Sukegawa, A., Sakurai, S., Masaki, K., et al., “Safety Design of Radiation Shielding for JT-60SA,” Fusion Eng. Des., 82, 2799 (2007). 144) Takado, N., Matsushita, D., (Inoue, T.), et al., “Effect of Energy Relaxation of H0 Atoms at the Wall on the Production Profile of H- Ions in Large Negative Ion Sources,” Rev. Sci. Instrum., 79, 02A503 (2008). 145) Takahashi, K., Kobayashi, N., Omori, J., et al., “Progress on Design and Development of ITER Equatorial Launcher; Analytical Investigation and R&D of the Launcher Components for the Design Improvement,” Fusion Sci. Tech., 52, 266 (2007). 146) Takahashi, Y., Yoshida, K., Nabara, Y., et al., “Stability and Quench Analysis of Toroidal Field Coils for ITER,” IEEE Trans. Appl. Superconduct., 17, 2426 (2007). 147) Takeda, N., Kakudate, S., Nakahira, M., et al., “Performance Test of Diamond-Like Carbon Films for Lubricating ITER Blanket Maintenance Equipment Under GPa-Level High Contact Stress,” Plasma Fusion Res., (Internet) 2, 052 (2007). 148) Takenaga, H., and JT-60 Team, “Overview of JT-60U Results for the Development of a Steady-State Advanced Tokamak Scenario,” Nucl. Fusion, 47, S563 (2007). 149) Takenaga, H., Kubo, H., Sueoka, M., et al., “Response of Fusion Gain to Density in Burning Plasma Simulation on JT-60U,” Nucl. Fusion, 48, 035011 (2008). 150) Takenaga, H., Oyama, N., Isayama, A., et al., “Observation on Decoupling of Electron Heat Transport and Long-Spatial-Scale Density Fluctuations in a JT-60U Reversed Shear Plasma,” Plasma Phys. Control. Fusion, 49, 525 (2007). 151) Takizuka, T., Oyama, N., Hosokawa, M., “Effect of Radial Transport Loss on the Asymmetry of ELM Heat Flux,” Contrib. Plasma Phys., 48, 207 (2008). 152) Tamai, H., Fujita, T., Kikuchi, M., et al., “Prospective Performances in JT-60SA Towards the ITER and DEMO Relevant Plasmas,” Fusion Eng. Des., 82, 541 (2007). 153) To, K., Shikama, T., (Yamauchi, M.), et al., “Search for Luminescent Materials Under 14 MeV Neutron Irradiation,” J. Nucl. Mater., 367-370, 1128 (2007). 154) Tobari, H., Seki, T., Takado, N., et al., “Negative Ion Production in High Electron Temperature Plasmas,” Plasma Fusion Res., (Internet) 2, 022 (2007). 155) Tobita, K., Nishio, S., Sato, M., et al., “SlimCS; Compact Low Aspect Ratio DEMO Reactor with Reduced-Size Central Solenoid,” Nucl. Fusion, 47, 892 (2007). 156) Tomita, Y., Smirnov, R.D., Takizuka, T., et al., “Effect of Oblieque Magnetic Field on Release Conditions of Dust Particle from Plasma-Facing Wall,” Contrib. Plasma Phys., 48, 285 (2008). 157) Tomita, Y., Smirnov, R., (Takizuka, T.), et al., “Effect of Truncation of Electron Velocity Distribution on Release of Dust Particle from Plasma-Facing Wall,” J. Nucl. Mater., 363-365, 264 (2007). 158) Tsuchiya, K., Kizu, K., Ando, T., et al., “Design of the Superconducting Coil System in JT-60SA,” Fusion Eng. Des., 82, 1519 (2007). 159) Tsuchiya, K., Hoshino, T., Kawamura, H., et al., “Development of Advanced Tritium Breeders and Neutron Multipliers for DEMO Solid Breeder Blankets,” Nucl. Fusion, 47, 1300 (2007). 160) Tsuchiya, K., Kawamura, H., Ishida, T., “Compatibility Between Be-Ti Alloys and F82H Steel,” J. Nucl. Mater., 367-370, 1018 (2007). 161) Tsuchiya, K., Kawamura, H., Ishida, T., “Effect of Ti Content on Compatibility Between Be-Ti and SS316LN,” Nuclear Technology 159, 228 (2007). 162) Tsuchiya, K., Shimizu, M., Kawamura, H., et al., “Effect of Re-Irradiation by Neutrons on Mechanical Properties of Un-Irradiated/Irradiated SS316LN Weldments,” J. Nucl. Mater., 373, 212 (2008). 163) Tsuru, D., Enoeda, M., Akiba, M., “Pressurizing Behavior on Ingress of Coolant Into Pebble Bed of Blanket of Fusion DEMO Reactor,” Fusion Eng. Des., 82, 2274 (2007). 164) Ueda, Y., Fukumoto, M., (Miya, N.), et al., “Surface Studies of Tungsten Erosion and Deposition in JT-60U,” J. Nucl. Mater., 363-365, 66 (2007). 165) Urano, H., and JT-60 Team, “H-mode Pedestal Structure in the Variation of Toroidal Rotation and Toroidal Field Ripple in JT-60U,” Nucl. Fusion, 47, 706 (2007). 166) Wakai, E., Ando, M., Sawai, T., et al., “Effect of Helium and Hydrogen Production on Irradiation Hardening of F82H Steel Irradiated by Ion Beams,” Materials Transactions 48, 1427 (2007). 167) Wakai, E., Ando, M., Sawai, T., et al., “Effect of Heat Treatments on Tensile Properties of F82H Steel Irradiated by Neutrons,” J. Nucl. Mater., 367-370, 74 (2007). 168) Watanabe, H., Nakamura, N., (Nakano, T.), et al., “X-ray Spectra from Neon-Like Tungsten Ions in the Interaction with Electrons,” Plasma Fusion Res., (Internet) 2, 027 (2007). 169) Yamada, H., Takenaga, H., Suzuki, T., et al., “Density Limit in Discharges with High Internal Inductance on JT-60U,” Nucl. Fusion, 47, 1418 (2007). 170) Yamaguchi, T., Kawano, Y., Kusama, Y., “Sensitivity Study for the Optimization of the Viewing Chord Arrangement of the ITER Poloidal Polarimeter,” Plasma Fusion Res., (Internet) 2, S1112 (2007). 171) Yamamoto, I., Nishitani, T., Sagara, A., “Overview of Recent Japanese Activities and Plans in Fusion Technology,” Fusion Sci. Tech., 52, 347 (2007). 172) Yamamoto, Y., Yamanishi, T., (Isobe, K.), et al., “Fundamental Study on Purity Control of the Liquid Metal Blanket Using Solid Electrolyte Cell,” Fusion Sci. Tech., 52, 692 (2007). 173) Yamauchi, M., Nishitani, T., Nishio, S., et al., “Activation Analysis for Sequential Reactions of a Fusion Demo-Reactor,” Fusion Sci. Tech., 52, 781 (2007). 174) Yoshida, M., and JT-60 Team, “Role of Pressure Gradient on Intrinsic Toroidal Rotation in Tokamak Plasmas,” Physical Review Letters, 100, 105002 (2008). 175) Yoshida, M., and JT-60 Team, “Momentum Transport and Plasma Rotation Profile in Toroidal Direction in JT-60U L-Mode Plasmas,” Nucl. Fusion, 47, 856 (2007). 176) Zanino, R., Astrov, M., (Takahashi, Y.), et al., “Predictive Analysis of the ITER Poloidal Field Conductor Insert (PFCI) Test Program,” IEEE Trans. Appl. Superconduct., 17, 1353 (2007). 177) Zucchetti, M., El-Guebaly, L.A., (Tobita, K.), et al., “The Feasibility of Recycling and Clearance of Active Materials from Fusion Power Plants,” J. Nucl. Mater., 367-370, 1355 (2007). A.1.3 List of Papers Published in Conference Proceedings 1) Hayashi, T., Sakurai, S., Masaki, K., et al., “Conceptual Design of Divertor Cassette Handling by Remote Handling System for JT-60SA,” Proc. 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 (2007). 2) Isayama, A., and JT-60 Team, “Control of Current Profile and Instability by Radiofrequency Wave Injection in JT-60U and Its Applicability in JT-60SA,” AIP Conference Proceedings 933, 229 (2007). 3) Isobe, K., Uzawa, M., Yamanishi, T., et al., “Development of Ceramic Electrolysis Method for Processing High-Level Tritiated Water,” STI/PUB/1284 (CD-ROM), 7 (2007). 4) Kobayashi, N., Bigelow, T., Bonicelli, T., et al., “Design of Electron Cyclotron Heating and Current Drive System of ITER,” AIP Conference Proceedings 933, 413 (2007). 5) Koide, Y., “Development in Diagnostics Application to Control of Advanced Tokamak Plasma,” AIP Conference Proceedings 988, 413 (2008). 6) Kondoh, T., Kawano, Y., Hatae, T., et al., “Progress in Development of Collective Thomson Scattering Diagnostic with High Power CO2 Laser,” NIFS-PROC-68, 126 (2007). 7) Matsunami, N., Nakano, T., “Current Status of Chemical Sputtering of Graphite and Related Materials,” Proc. International Symposium on EcoTopia Science 2007 (ISETS '07) (CD-ROM), 321 (2007). 8) Nishitani, T., Ishikawa, M., Kondoh, T., et al., “Absolute Neutron Emission Measurement in Burning Plasma Experiments,” AIP Conference Proceedings 988, 267 (2008). 9) Sugie, T., Ogawa, H., Kasai, S., et al., “Spectroscopic Measurement System for ITER Divertor Plasma; Impurity Influx Monitor (Divertor),” AIP Conference Proceedings 988, 218 (2008). 10) Sukegawa, A., Oikawa, A., Miya, N., et al., “Estimation of Low Level Waste by a Regulatory Clearance in JT-60U Fusion Device,” Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 (2007). 11) Takahashi, K., Kajiwara, K., Kasugai, A., et al., “Investigation of Transmission Characteristic in Corrugated Waveguide Transmission Lines for Fusion Application,” Proc. 8th IEEE International Vacuum Electronics Conference (IVEC 2007), 275 (2007). 12) Takato, N., Hanatani, J., (Inoue, T.), et al., “Numerical Analysis of the Hydrogen Atom Density in a Negative Ion Source,” AIP Conference Proceedings 925, 38 (2007). 13) Takenaga, H., Oyama, N., Asakura, N., et al., “Divertor Density Measurements Using mm-Wave Interferometer in JT-60U,” NIFS-PROC-68, 62 (2007). A.1.4 List of other papers 1) Ando, M., Tanigawa, H., Shiba, K., et al., “Irradiation Creep Behavior of Reduced Activation Ferritic/Martensitic Steel Irradiated in HFIR,” J. Jpn. Inst. Met., 71, 559 (2007) (in Japanese). 2) Ando, M., Wakai, E., Okubo, N., et al., “Extra-Irradiation Hardening of Reduced Activation Ferritic/Martensitic Steel by Multi-Ion Irradiation,” J. Jpn. Inst. Met., 71, 1107 (2007) (in Japanese). 3) Ando, T., “Information on ITER Project, 3,” J. Plasma Fusion Res., 82, 533 (2007) (in Japanese). 4) Ando, T., “Information on ITER Project, 5,” J. Plasma Fusion Res., 83, 775 (2007) (in Japanese). 5) Ando, T., “Information on ITER Project, 7,” J. Plasma Fusion Res., 84, 69 (2008) (in Japanese). 6) Ando, T., “Information on ITER Project, 8,” J. Plasma Fusion Res., 84, 164 (2008) (in Japanese). 7) Araki, M., Kamada, Y., Mori, M., et al., “Current Status and Evolution of Nuclear Fusion Development; Aiming at the Achievement of the Nuclear Fusion Energy,” Electrical Review, 92, 38 (2007) (in Japanese). 8) Asakura, N., “Fast Plasma Flow in Tokamak Divertor and Scrape-Off Layer,” J. Plasma Fusion Res., 83, 501 (2007) (in Japanese). 9) Fujita, T., Fukuda, T., (Shinohara, K.), et al., “Report on ITPA (International Tokamak Physics Activity) Meeting, 20,” J. Plasma Fusion Res., 84, 70 (2008) (in Japanese). 10) Hayashi, T., Sakurai, S., Masaki, K., et al.,”Conceptual Design Study of Exchange of the In-Vessel Components by Remote Handling System for JT-60SA,” Transactions of the American Nuclear Society, 96, 783 (2007). 11) Higuchi, M., “Systematization of Standards,” Maintenology, 6, 73 (2007) (in Japanese). 12) Ida, M., Idomura, Y., Tokuda, S., “Gyrokinetic Simulation of Fusion Plasma Turbulence, 1; Modelling and Numerical Methods,” Papers in the 2007 Annual Meeting of The Japan Society of Fluid Mechanics (CD-ROM), 5 (2007) (in Japanese). 13) Ide, S., “Present Status of Tokamak Research Towards Steady-State Operation,” J. Plasma Fusion Res., 83, 415 (2007) (in Japanese). 14) Ikeda, Y., “Recent Progress of Fusion Research,” Text of The 39th Summer Seminar concerning the Reactor Physics by Atomic Energy Society of Japan, 160 (2007) (in Japanese). 15) Koide, Y., “On the JT-60 Joint Research Prize,” J. Plasma Fusion Res., 84, 230 (2008) (in Japanese). 16) Koizumi, N., Nishimura, A., “Intelligible Seminar on Fusion Reactors, 9; Superconducting Coil to Generate Magnetic Field for Plasma Confinement,” J. Nucl. Sci. Tech., 47, 703 (2007) (in Japanese). 17) Kurihara, K., “Principles of Magnetic Measurements,” Handbook of the Modern Electric Power, 129 (2007) (in Japanese). 18) Maegawa, T., Yoshimatsu, K., Sato, S., “14 MeV Neutron Irradiation Test of Low-Activation Concrete with Limestone Powder,” Denryoku Doboku, 86 (2008) (in Japanese). 19) Matsui, K., Koizumi, N., Nabara, Y., et al., “Evaluation of Thermal Strain Caused by Nb3Sn Reaction Heat Treatment for the ITER Cable-in-Conduit Conductors,” J. Cryo. Soc. Jpn., 42, 311 (2007) (in Japanese). 20) Mori, M., “ITER Project in Detail,” Genshiryoku eye, 53, 11 (2007) (in Japanese). 21) Mori, M., Okuno, K., Sakamoto, K., “Information on ITER Project, 4,” J. Plasma Fusion Res., 83, 643 (2007) (in Japanese). 22) Mori, M., Yoshida, H., “Information on ITER Project, 6,” J. Plasma Fusion Res., 83, 928 (2007) (in Japanese). 23) Ninomiya, H., “Nuclear Fusion,” Thermal Power Generation and Atomic Power Generation, 58, 1038 (2007) (in Japanese). 24) Nishitani, T., “Detail of the Broader Approach Project,” Genshiryoku eye, 53, 21 (2007) (in Japanese). 25) Ohira, S., “Information of Broader Approach Activities, 2,” J. Plasma Fusion Res., 83, 398 (2007) (in Japanese). 26) Ohira, S., “Information of Broader Approach Activity, 3,” J. Plasma Fusion Res., 83, 587 (2007) (in Japanese). 27) Ohira, S., “Information of Broader Approach Activity, 4,” J. Plasma Fusion Res., 83, 716 (2007) (in Japanese). 28) Ohira, S., “Information of Broader Approach Activity, 5,” J. Plasma Fusion Res., 83, 853 (2007) (in Japanese). 29) Ohira, S., “Information of Broader Approach Activity, 7,” J. Plasma Fusion Res., 84, 229 (2008) (in Japanese). 30) Okuno, K., “Recent Progress in the Technology Development for the ITER Superconducting Coils,” Superconductivity Communications, 16, 6 (2007) (in Japanese). 31) Ozeki, T., “Realization of a Nuclear Fusion Experiment from a Remote Place by the Advanced Security,” RIST News, 3 (2007) (in Japanese). 32) Ozeki, T., “Remote Hierarchical Environment of Experiment and Simulation,” Simulation, 27, 27 (2008) (in Japanese). 33) Ozeki, T., Watanabe, K., “Avoidance and Suppression of MHD Instability,” J. Plasma Fusion Res., 83, 44 (2007) (in Japanese). 34) Sakamoto, Y., Ida, K., “Control of High Confinement Plasmas,” J. Plasma Fusion Res., 83, 439 (2007) (in Japanese). 35) Sasao, M., Kusama, Y., Kawano, Y., et al., “Report of Meetings of ITPA (International Tokamak Physics Activity), 19,” J. Plasma Fusion Res., 83, 779 (2007) (in Japanese). 36) Sawahata, A., Tanigawa, H., Enomoto, M., “Effects of Electro-Slag Remelting on Inclusion Formation and Impact Property of Reduced Activation Ferritic/Martensitic Steels,” J. Jpn. Inst. Met., 72, 176 (2008) (in Japanese). 37) Shinohara, K., Takechi, M., Ishikawa, M., “Interaction Between Alfvén Eingenmodes and Energetic Ions on Tokamaks,” J. Plasma Fusion Res., 83, 873 (2007) (in Japanese). 38) Suzuki, S., Enoeda, M., Matsuda, H., et al., “Numerical Evaluation of Crack Propagation of ITER First Wall with an Initial Interfacial Defect,” Transactions of the Atomic Energy Society of Japan, 6, 365 (2007) (in Japanese). 39) Suzuki, T., “Current profile control in advanced tokamak plasmas,” J. Plasma Fusion Res., 83, 434 (2007) (in Japanese). 40) Takahashi, Y., “How is the ITER Project Going on?,” Chodendo Web 21 (Internet), 25 (2008) (in Japanese). 41) Takeda, N., Kakudate, S., Nakahira, M., et al., “Current Status of Research and Development on Remote Maintenance for Fusion Components,” J. Plasma Fusion Res., 84, 100 (2008) (in Japanese). 42) Takemura, M., “My First stay in Provence in the South of France,” FAPIG, 3 (2007) (in Japanese). 43) Takenaga, H., Morisaki, T., “Fuelling and Heat / Particle Control,” J. Plasma Fusion Res., 83, 453 (2007) (in Japanese). 44) Tobita, K., “Challenges in Technology for Attractive Fusion Energy,” J. Inst. Electr. Eng. Jpn., 128, 86 (2008) (in Japanese). 45) Urano, H., “Sun on the Earth,” Enjinia Ring, 9 (2007) (in Japanese). 46) Urano, H., Matsumoto, T., Matsunaga, G., “Conference Report on "10th Plasma Research Workshop by Young Scientists" J. Plasma Fusion Res., 82, 534 (2007) (in Japanese). 47) Ushigusa, K., Seki, M., Ninomiya, H., et al., “Research and Development of Nuclear Fusion,” Genshiryoku Handbook, 906 (2007) (in Japanese). 48) Yamanishi, T., Iwai, Y., Isobe, K., et al., “Developments of Water Detritiation Systems in a Fusion Reactor,” J. Plasma Fusion Res., 83, 545 (2007) (in Japanese). 49) Yoshida, H., “The Trend of Fusion Energy Research and Development in the Participant Party and Countries in the ITER Project,” Genshiryoku eye, 53, 31 (2007) (in Japanese). A.2 Organization of Fusion Research and Development Directorate A.3 Personnel Data A.3.1 Staff in Fusion Research and Development Directorate of JAEA Fusion Research and Development Directorate TSUNEMATSU Toshihide (Director General) NINOMIYA Hiromasa (Deputy Director General) OKUMURA Yoshikazu (Deputy Director General) NAGAMI Masayuki (Supreme Researcher) TANI Keiji (Senior Principal Researcher) YOSHIDA Hidetoshi (Principal Researcher) SEKI Masahiro (Invited Researcher) SEKI Shogo (Invited Researcher) SHIMOMURA Yasuo (Invited Researcher) MATSUI Hideki (Invited Researcher) KOHYAMA Akira (Invited Researcher) IDA Katsumi (Invited Researcher) KISHIMOTO Yasuaki (Invited Researcher) Research and Development Co-ordination and Promotion Office NINOMIYA Hiromasa (General Manager) WATANABE Tsutomu (Deputy General Manager) GUNJI Masato Hayashi Kazuyuki KIMURA Shoko KOYANAGI Daisaku MATSUMOTO Hiroyuki MORITA Hisao NEMOTO Tetsuo ONOZAKI Kazutoyo SOGA Yoriko TSUCHIDA Tatsuro YOSHIDA Hiroshi ITER Project Promotion Group SHIRAI Hiroshi (Group Leader) DOI Kenshin OGAWA Toshihide Fusion Research Coordination Group USHIGUSA Kenkichi (Group Leader) ISEI Nobuaki OOHARA Hiroshi Division of Advanced Plasma Research KIKUCHI Mitsuru (Unit Manager) OZEKI Takahisa (Senior Principal Researcher) JT-60 Advanced Program Group KAMADA Yutaka (Group Leader) TAKENAGA Hidenobu URANO Hajime JT-60SA Design Integration Group MATSUKAWA Makoto (Group Leader) YOSHIDA Kiyoshi (Deputy Group Leader) KAWASHIMA㻃 Hisato TSUCHIYA Katsuhiko KIZU Kaname TSUKAHARA Yoshimitsu HOSHI Ryo (*4) MATSUMURA Hiroshi (*4) SUZUKI Yutaka (*17) EDAYA Masahiro (*31) SATO Fujio (*7) KOMEDA Masao (*16) Collaborative Research Group KOIDE Yoshihiko (Group Leader) IDE Shunsuke (Deputy Group Leader) HOSHINO Katsumichi KAMIYA Kensaku KONOSHIMA Shigeru (*1) Tokamak Analysis Group OZEKI Takahisa (Group Leader) NAITO Osamu (Deputy Group Leader) AIBA Nobuyuki (*23) HAMAMATSU Kiyotaka HAYASHI Nobuhiko HONDA Mitsuru (*23) KIYONO Kimihiro KAMATA Isao (*24) KOMINATO Toshiharu (*2) SAKATA Shinya SATO Minoru SHIMIZU Katsuhiro SUZUKI Mitsuhiro (*31) TAKIZUKA Tomonori (*1) Tokamak Experimental Group ITAMI Kiyoshi (Group Leader) ASAKURA Nobuyuki (Deputy Group Leader) CHIBA Shinichi FUJIMOTO Kayoko (*23) HANAWA Osamu (*22) HATAE Takaki ISAYAMA Akihiko KOIKE Tomonori (*22) KITAMURA Shigeru KOJIMA Atsushi (*23) MATSUNAGA Go MIYAMOTO Atsushi (*21) NAKANO Tomohide NUMATA Hiroyuki (*22) OYAMA Naoyuki SAKAMOTO Yoshiteru SHINOHARA Kouji SUNAOSHI Hidenori SUZUKI Takahiro TSUBOTA Naoaki (*21) TSUTSUMI Kazuyoshi (*21) UEHARA Kazuya (*1) YOSHIDA Maiko (*25) Plasma Theory & Simulation Group TOKUDA Shinji (Group Leader) IDOMURA Yasuhiro ISHII Yasutomo KAGEI Yasuhiro (*25) LESUR Maxime (*3) MATSUMOTO Taro MIYATO Naoaki SUZUKI Yoshio TUDA Takashi (*1) Fusion Reactor Design Group TOBITA Kenji (Group Leader) NAKAMURA Yukiharu KURITA Gen-ichi NISHIO Satoshi JT-60SA Torus Development Group SAKASAI Akira (Group Leader) HIGASHIJIMA Satoru KASHIWA Yoshitoshi MASAKI Kei SAKURAI Shinji SHIBAMA Yusuke TAKECHI Manabu Division of Tokamak System Technology HOSOGANE Nobuyuki (Unit Manager) YAMAMOTO Takumi (Senior Principal Researcher) FUJII Tsuneyuki (Senior Principal Researcher) Tokamak Control Group KURIHARA Kenichi (Group Leader) AKASAKA Hiromi HOSOYAMA Hiromi (*8) KAMIKAWA Masaaki (*10) KAWAMATA Youichi MATSUKAWA Tatsuya (*21) OHMORI Yoshikazu OKANO Jun SATO Tomoki (*31) SEIMIYA Munetaka (*1) SHIBATA Kazuyuki (*22) SHIMADA Katsuhiro SUEOKA Michiharu TERAKADO Hiroyuki (*7) TERAKADO Tsunehisa TOTSUKA Toshiyuki YAMAUCHI Kunihito WADA Kazuhiko (*22) Tokamak Device Group SATO Masayasu (Group Leader) ARAI Takashi HAGA Saburo (*21) ICHIGE Hisashi ISHIGE Youichi (*22) KAMINAGA Atsushi MATSUZAWA Yukihiro (*21) MIYO Yasuhiko NISHIYAMA Tomokazu OKANO Fuminori SASAJIMA Tadayuki SATO Yoji (*21) SUZUKI Yozo (*4) YAGYU Jun-ichi RF Heating Group HOSOGANE Nobuyuki (Group Leader) HIRANAI Shinichi HASEGAWA Koichi IGARASHI Koichi (*21) KOBAYASHI Takayuki (*23) MORIYAMA Shinichi SATO Fumiaki (*21) SAWAHATA Masayuki SHIMONO Mitsugu SUZUKI Sadaaki SUZUKI Takashi (*22) TERAKADO Masayuki WADA Kenji (*21) YOKOKURA Kenji NBI Heating Group IKEDA Yoshitaka (Group Leader) AKINO Noboru EBISAWA Noboru HANADA Masaya HONDA Atsushi KAMADA Masaki (*23) KAWAI Mikito KAZAWA Minoru KIKUCHI Katsumi (*22) KOBAYASHI Kaoru (*4) KOMATA Masao MOGAKI Kazuhiko NOTO Katsuya (*21) OOASA Kazumi OSHIMA Katsumi (*21) SASAKI Shunichi SHIMIZU Tastuo (*21) SHINOZAKI Shinichi TAKENOUCHI Tadashi (*29) TANAI Yutaka (*22) USUI Katsutomi Safety Assessment Group MIYA Naoyuki (Group Leader) SUKEGAWA Atsuhiko HAYASHI Takao OIKAWA Akira (*1) Division of Fusion Energy Technology TAKATSU Hideyuki (Unit Manager) NISHITANI Takeo (Senior Principal Researcher) Blanket Technology Group AKIBA Masato (Group Leader) ENOEDA Mikio EZATO Koichiro HIROSE Takanori HOMMA Takashi (*1) MOHRI Kensuke (*12) NISHI Hiroshi SEKI Yohji (*23) SUZUKI Satoshi TANIGAWA Hisashi TANZAWA Sadamitsu TSURU Daigo YOKOYAMA Kenji Plasma Heating Group SAKAMOTO Keishi (Group Leader) INOUE Takashi (Deputy Group Leader) DAIRAKU Masayuki IKEDA Yukiharu KASHIWAGI Mieko KASUGAI Atsushi KAJIWARA Ken (*25) KOBAYASHI Noriyuki (*30) KOMORI Shinji (*22) OMINE Takashi (*31) TAKAHASHI Koji TAKESHIMA Yuichirou (*5) TANIGUCHI Masaki TOBARI Naoyuki (*25) UMEDA Naotaka WATANABE Kazuhiro YAMAMOTO Masanori (*4) Tritium Technology Group YAMANISHI Toshihiko (Group Leader) ARITA Tadaaki(*26) HAYASHI Takumi HOSHI Shuichi(*22) INOMIYA Hiroshi(*22) ISOBE Kanetsugu IWAI Yasunori KAWAMURA Yoshinori KOBAYASHI Kazuhiro NAKAMURA Hirofumi SHU Wataru SUZUKI Takumi YAMADA Masayuki Fusion Structural Materials Development Group TAKATSU Hideyuki (Group Leader) TANIGAWA Hiroyasu (Deputy Group Leader) ANDO Masami NOZAWA Takashi (*25) OGIWARA Hiroyuki (*23) SAWAHATA Atsushi (*5) NAKATA Toshiya (*18) Fusion Neutronics Group KONNO Chikara (Group Leader)) ABE Yuichi IIDA Hiromasa (*1) KAWABE Masaru (*22) KUTSUKAKE Chuzo OCHIAI Kentaro OKADA Koichi (*28) OHNISHI Seiki (*25) SATO Satoshi TAKAKURA Kosuke (*21) TANAKA Shigeru Blanket Irradiation and Analysis Group HAYASHI Kimio (Group Leader) NAKAMICHI Masaru HOSHINO Tsuyoshi YONEHARA Kazuo (*9) HASEGAWA Teiji (*22) NAMEKAWA yoji (*22) TSUCHIYA Kunihiko Division of ITER Project YOSHINO Ryuji (Unit Manager) ITER Project Management Group KOIZUMI Koichi (Group Leader) NEYATANI Yuzuru (Deputy Group Leader) KITAZAWA Siniti IWAMA Yasushi (*21) SENGOKU Akio (*21) KOIKE Kazuhisa (*21) YAGUCHI Eiji (*13) SATO Koichi (*31) SATO Kazuyoshi HIGUCHI Masahisa (*27) ITER International Coordination Group MORI Masahiro (Group Leader) ANDO Toshiro (Deputy Group Leader) OIKAWA Toshihiro ODAJIMA Kazuo (*1) ITER Tokamak Device Group SHIBANUMA Kiyoshi (Group Leader) KAKUDATE Satoshi KOZAKA Hiroshi (*21) MATSUMOTO Yasuhiro (*30) TAGUCHI Kou (*22) TAKEDA Nobukazu ITER Superconducting Magnet Technology Group OKUNO Kiyoshi (Group Leader) NAKAJIMA Hideo (Deputy Group Leader) HAMADA Kazuya HEMMI Tsutomu ISONO Takaaki KAWANO Katsumi KOIZUMI Norikiyo MATSUI Kunihiro NABARA Yoshihiro NAKAHIRA Masataka NIIMI Kenichiro (*11) NUNOYA Yoshihiko OHMORI Junji (*30) OKUI Yoshio (*14) OSHIKIRI Masayuki (*22) SEO Kazutaka (*19) SHIMANE Hideo (*13) SHIMIZU Tatuya (*13) TAKAHASHI Yoshikazu TAKANO Katsutoshi (*22) TSUTSUMI Fumiaki (*31) ITER Diagnostics Group KUSAMA Yoshinori (Group Leader) FUJIEDA Hirobumi HAYASHI Toshimitsu (*20) ISHIKAWA Masao (*25) KAJITA Shin (*25) KAWANO Yasunori KONDOH Takashi OGAWA Hiroaki SUGIE Tatsuo TAKEI Nahoko (*23) YAMAGUCHI Taiki (*23) YAMAMOTO Tsuyoshi (*8) Division of Rokkasho BA Project OKUMURA Yoshikazu (Unit Manager) BA Project Promotion Group OHIRA Shigeru (Group Leader) EJIRI Shintaro (Deputy Group Leader) BA Project Coordination Group OHIRA Shigeru (Group Leader) TAKEMOTO Junpei (*10) IFMIF Development Group OKUMURA Yoshikazu (Group Leader) NAKAMURA Hiroo (Deputy Group Leader) ASAHARA Hiroo(*21) IDA Mizuho (*6) MAEBARA Sunao KIKUCHI Takayuki KOJIMA Toshiyuki(*21) KUBO Takashi (*15) MIYASHITA Makoto (*30) YONEMOTO Kazuhiro(*10) A.3.2 Staff in ITER Organization and Project Teams of the Broader Approach Activities ITER Organization Office of Director-General MATSUMOTO Hiroshi (Head) Project Office TADA Eisuke (Head) Department for Administration IDE Toshiyuki Department for Fusion Science & Technology SHIMADA Michiya Department for Central Engineering & Plant Support MARUYAMA So Project Teams of the Broader Approach Activities International Fusion Energy Research Center ARAKI Masanori (Project Leader) Satellite Tokamak Programme ISHIDA Shinichi (Project Leader) FUJITA Takaaki KIMURA Haruyuki OSHIMA Takayuki SEKI Masami TAMAI Hiroshi IFMIF EVEDA SUGIMOTO Masayoshi (Deputy Project Leader) NAKAMURA Kazuyuki SHINTO Katsuhiro(*25) A.3.3 Collaborating Laboratories Tokai Research and Development Center Nuclear Science and Engineering Directorate Irradiation Field Materials Research Group JITSUKAWA Shiro (Group Leader) FUJII Kimio OKUBO Nariaki TANIFUJI Takaaki WAKAI Eiichi YAMAKI Daiju Research Group for Corrosion Damage Mechanism MIWA Yukio Quantum Beam Science Directorate Nanomaterials Synthesis Group TAGUCHI Tomitsugu Oarai Research and Development Center Technology Development Department MIYAKE Osamu (Deputy Director) Advanced Liquid Metal Technology Experiment Section YOSHIDA Eiichi (General Manager) HIRAKAWA Yasushi Advanced Nuclear System Research and Development Directorate Innovative Technology Group ARA Kuniaski (Group Leader) OTAKE Masahiko *1 Contract Staff *2 Customer System Co., Ltd. *3 Ecole Polytechnique (France) *4 Hitachi, Ltd. *5 Ibaraki University *6 Ishikawajima-Harima Heavy Industries Co., Ltd. *7 JP HYTEC Co., Ltd. *8 Japan EXpert Clone Corp. *9 KAKEN Co., Ltd. *10 Kandenko Co., Ltd. *11 Kawasaki Heavy Industries, Ltd. *12 Kawasaki Plant Systems, Ltd. *13 KCS Corporation *14 Kobe Steel, Ltd. *15 Kumagai Gumi Co., Ltd. *16 MAYEKAWA MFG. CO., LTD *17 Mitsubishi Heavy Industries, Ltd. *18 Muroran Institute of Technology *19 National Institute for Fusion Science *20 NEC Corporation *21 Nippon Advanced Technology Co., Ltd. *22 Nuclear Engineering Co., Ltd. *23 Post-Doctoral Fellow *25 Princeton Plasma Physics Laboratory (USA) *24 Research Organization for Information Science & Technology *25 Senior Post-Doctoral Fellow *26 Sumitomo Heavy Industries, Ltd. *27 The Japan Atomic Power Company *28 Tohoku University *29 Tomoe Shokai Co., Ltd. *30 Toshiba Corporation *31 Total Support Systems